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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

Commission (the Commission on 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

[[Page 11126]] 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 3, 1995, through February 16, 1995.

The last biweekly notice was published on February 15, 1995 (60 FR

8739).

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 March 31, 1995, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no [[Page 11127]] 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 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 section, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: December 7, 1994.

Description of amendment requests: The proposed amendment would

revise the capacity of the ultimate heat sink (UHS) as described in the

bases of Technical Specification 3/4.7.5, ``Ultimate Heat Sink,'' from

providing a 27-day cooling water supply to providing a 26-day cooling

water supply. In addition, the reference to Regulatory Guide 1.27 in

the bases of this TS would also be revised to reference the January

1976 revision rather than the March 1974 revision.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1--Does the proposed change involve a significant

increase in the probability or consequences of an accident

previously evaluated?

The Essential spray pond system and the UHS do not initiate any

accidents in Chapters 6 or 15 of the UFSAR [Updated Final Safety

Analysis Report]. The justification and basis for the time that the

UHS is available is not changed and continues to be consistent with

the guidance in Regulatory Guide 1.27. The existing Technical

Specification requirements and those components to which they apply

are not altered by this Technical Specification amendment.

Therefore, the change to the bases for Technical Specification 3/

4.7.5 does not increase the probability of occurrence or the

consequences of any previously evaluated accident.

Standard 2--Does the proposed change create the possibility of a

new or different kind of accident from any accident previously

evaluated?

The requirements for Technical Specification 3/4.7.5 are not

changed. This amendment has no impact on plant maintenance, testing,

shutdown equipment, or component qualification. Therefore, the

possibility of a new or different kind of accident is not created by

this amendment.

Standard 3--Does the proposed change involve a significant

reduction in a margin of safety?

The change to the bases for Technical Specification 3/4.7.5 does

not significantly alter existing Technical Specification

requirements or those coponments to which they apply. The

justification and basis for the time that the UHS is available

without makeup is not changed and continues to be consistent with

the guidance in Regulatory Guide 1.27. Regulatory Guide 1.27 states

that ``A capacity less than 30 days may be acceptable if it can be

demonstrated that replenishment can be effected to ensure that

continuous capability of the sink to perform its safety functions,

taking into account the availability of replenishment equipment and

limitations that may be imposed on ``freedom of movement'' following

an accident.'' This change does not effect the continuous capability

of the UHS to perform its safety function of providing decay heat

removal capability following an accident. The change updates the

design basis of the UHS using more realistic conditions based on

plant experience. Therefore, the change in the capacity of the UHS

without makeup from 27 days to 26 days will not involve a

significant reduction in margin of safety for the ultimate heat

sink.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

Attorney for licensees: Nancy C. Loftin, Esq., Corporation

Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999,

Mail Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: Theodore R. Quay.

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

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

Date of amendment request: June 18, 1992, as supplemented December

8, 1992, and revised February 3, 1995.

Description of amendment request: The proposed Technical

Specification (TS) amendment adds limiting conditions of operation and

surveillance requirements for the pressurizer power-operated relief

valves (PORVs) and their associated block valves whenever average

temperature (Tavg) is above 350 degrees F or the reactor is critical.

Specifications have also been added for low-temperature overpressure

protection whenever Tavg is less than 350 degrees F and the reactor

coolant system is not vented to the containment. The February 3, 1995,

revision made editorial changes to previous TS pages and made changes

to conform with an additional provision of the guidance for

surveillance testing of the block valves associated with the

pressurizer PORVs. In addition, the licensee has requested an editorial

change to TS page 3.1.-11 to revise the references to two figures that

have been superseded.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The requested revision does not involve a significant

increase in the probability or consequences of an accident

previously [[Page 11128]] evaluated. The proposed revision to our

previous Technical Specification (TS) change request dated June 18,

1992, would help assure the availability of the block valves for

accident mitigation. The availability of the block valves for

accident mitigation has been found to outweigh any negative safety

consequences associated with full cycle testing of a block valve

isolating a pressurizer power-operated relief valves (PORV) with

``excessive'' seat leakage. There would be no significant increase

in the probability or consequences of an accident previously

evaluated since this event is fully bounded by the failing open of a

single pressurizer code safety relief valve event which is analyzed

in Chapter 15 of the Updated Final Safety Analysis Report.

Accordingly, the requested revision will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The requested revision to our previous TS change request does

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

from any accident previously evaluated. Periodic testing of the

block valves in accordance with the requested revision is only

intended to assure the functioning and capability of the block

valves. The requested revision will only clarify the conditions when

block valve surveillance testing is required. The performance of

this testing is intended to improve block valve availability and

thereby assure the capability of certain accident mitigation

strategies identified within Abnormal and Emergency Operating

Procedures. Therefore, the requested revision will not create the

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

previously evaluated.

3. The requested revision to our previous TS change request does

not involve a significant reduction in the margin of safety. The

requested revision is intended to help assure block valve

availability to support certain accident mitigation strategies. This

additional assurance of block valve availability and functioning

increases the margin of safety. Accordingly, the requested revision

will not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

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

NRC Project Director: William H. Bateman.

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

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

Date of amendment request: December 14, 1994.

Description of amendment request: The proposed amendments would

revise technical specifications related to allowed outage times (AOT)

and surveillance test intervals (STI) for certain actuation

instrumentation in the reactor protection system (RPS), primary

containment isolation system (PCIS), emergency core cooling system

(ECCS), recirculation pump trip, reactor core isolation cooling (RCIC),

control rod withdrawal block, monitoring, and feedwater/main turbine

trip systems. These changes are generally consistent with General

Electric topical reports which have been reviewed and approved by the

NRC. The changes also include revising the Feedwater/Main Turbine Trip

LCO 3.3.8 action statement to achieve consistency with existing

instrumentation LCOs; deleting the surveillance of the APRM Neutron

Flux--High, Setdown functional unit in Operational Condition 1;

revising the applicability of the provisions of Specification 4.0.4 to

several Reactor Protection System and Control Rod Withdrawal Block

Instrumentation surveillance requirements; adding the requirement to

perform shiftly channel checks for applicable RPS, PCIS, ECCS, and RCIC

instrumentation channels equipped with master trip units; and other

changes to correct typographical errors and to delete cycle specific

footnotes which are no longer applicable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

It has been determined that the changes do not constitute a

Significant Hazards Consideration. Based on the criteria for

defining a significant hazards consideration established in 10 CFR

50.92, operation of LaSalle County Station Units 1 and 2 in

accordance with the proposed amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

a. The proposed changes increase the STI and AOT for actuation

instrumentation supporting RPS, ECCS, Isolation, CRBF, RCIC, ATWS-

RPT, EOC-RPT, Monitoring, and Feedwater/Main Turbine Trip System

Actuation functions. There are no changes in instrumentation

configuration and function, and no instrumentation setpoints are

changed. Because of this there is no change in the probability of

occurrence of an accident or the consequences of an accident or the

consequences of malfunction of equipment. With respect to the

probability of equipment malfunction, topical reports prepared by GE

demonstrate that there is a reduction in scram frequency for the

RPS, but in the case of the ECCS there is a small increase in the

unavailability of the water injection function. This increase in

unavailability was judged acceptable by GE. The NRC concurred with

this conclusion in its review and approval of the topical reports.

The proposed changes are consistent with the Safety Evaluation

Reports issued for the topical reports.

b. The changes proposed for the Feedwater/Main Turbine Trip LCO

action statements provide actions which are consistent with

presently existing instrumentation LCOs. The design and function of

the feedwater/main turbine trip instrumentation to trip the

feedwater pumps and the main turbine upon detection of a Level 8

event is not altered. The probability and/or consequences of this

moderate frequency transient are not increased.

c. The APRM Neutron Flux--High, Setdown scram setting provides

adequate thermal margin between the setpoint and the safety limits

for operation at low pressure and low flow during a plant startup.

This function remains in effect until the mode switch is placed in

the Run (Operational Condition 1) position, at which time it is

bypassed. Deleting the requirement for the surveillance of the APRM

Neutron Flux--High, Setdown functional unit in Operational Condition

1 is appropriate since its function is not applicable in this mode.

This deletion serves to achieve consistency between Technical

Specification Tables and the Bases section.

d. The changes associated with Specification 4.0.4 are

administrative in nature and are intended to provide the plant

operators with better guidance for its application. In cases where

complete surveillances cannot be achieved, such as during a plant

shutdown, then the required surveillances will be performed within

24 hours of entering the Mode or condition in which the surveillance

is required. The stabilization of the plant will be of primary

consideration. This change does not affect the evaluation for any

accident presented in Chapter 15 of the UFSAR. The APRM Fixed

Neutron Flux--High quarterly functional tests most of the APRM

channel equipment associated with the APRM Neutron Flux--High,

Setdown scram.

Additionally, the expected result of the functional tests

associated with the SRMs, IRMs, and APRMs is to demonstrate the

operability of the instrumentation. Therefore, 24 hours is a

reasonable time to permit the surveillances to be performed upon

entering the mode or condition in which the surveillance is

required.

e. The proposal to include the performance of channel checks as

requirements of technical specifications is administrative in

nature. Presently, channel checks performed for the applicable

analog instrumentation in reactor vessel water level applications is

controlled solely by procedure. Adding this

[[Page 11129]] requirement to the technical specifications provides

for the appropriate controls of the surveillances, above and beyond

that presently controlled by procedure.

f. The proposed administrative changes are offered to correct

typographical errors and delete cycle specific footnotes which are

no longer applicable. The nature of the changes precludes them from

impacting previously analyzed accidents.

The proposed changes therefore do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

accident from any accident previously evaluated because:

a. The proposed changes increase the STI and AOT for certain

actuation instrumentation in the RPS, ECCS, Isolation, CRBF, RCIC,

ATWS-RPT, EOC-RPT, Monitoring, and Feedwater/Main Turbine Trip

systems. There are no changes in instrumentation configuration and

function, and no instrumentation setpoints are changed.

b. The changes to the Feedwater/Main Turbine Trip LCO action

statements allow the plant operators a maximum degree of operational

flexibility, while maintaining the instrumentation and protection

needed for terminating the feedwater controller failure transient.

The single failure proof criterion of the level sensors is

maintained, and the logic of the protective instrumentation is not

compromised. The changes to the LCO action statements do not

constitute a change to the facility or its operation as described in

the Safety Analysis Report.

c. Deleting the requirement for surveilling the APRM Neutron

Flux--High, Setdown functional unit in Operating Condition 1 does

not degrade thermal margins. The margin accommodates the anticipated

maneuvers associated with plant power ascension. During a plant

shutdown, rod insertion maneuvers, recirculation flow reduction, and

xenon build-in all contribute to negative reactivity insertion which

precludes the degradation and violation of thermal margins. The

functions of the APRMs required to be OPERABLE in Operational

Condition 1 which are in effect remain to ensure that reactor core

thermal margins are not compromised.

d. The conduct of neutron instrument functional tests in the

plant mode or condition in which the trips are applicable eliminates

unnecessary testing during normal plant operations. The expected

result of the functional testing is to demonstrate the operability

of the instruments. The failure of any single instrument channel

will neither cause nor prevent either a reactor scram or a control

rod block.

e. Including the performance of channel checks for the

applicable analog instrumentation as part of the technical

specifications transfers control of the required surveillances from

procedure to the technical specifications, as appropriate. The

administrative nature of this change does not alter the functions,

setpoints, or configuration of the associated instrumentation.

f. The administrative nature of the changes prevents them from

affecting the functions, setpoints, or configuration of the

associated instrumentation from being affected by the changes.

The proposed changes do not create the possibility for an

accident or malfunction of a different type than any previously

evaluated in the UFSAR.

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

because:

a. Setpoints are based upon the drift occurring during an 18

month calibration interval. The bases in the Technical

Specifications either do not discuss STI, or state ``* * * one

channel may be inoperable for brief intervals to conduct required

surveillance.'' The proposed changes are bounded by the analyses of

the topical reports. These analyses, which were prepared by GE and

approved by the NRC, examined the effects of extending STI and AOT

and found that the proposed changes would not involve a significant

reduction in the margin of safety.

b. The proposed changes to the turbine trip LCO action

statements do not change any of the settings of the Level 8

setpoints. The single failure criteria of the multiple level sensors

which sense and detect the Level 8 setpoint remains intact. The LCO

maintains the requirement that no single instrument failure will

prevent the feedwater pump turbines and main turbine trip on a valid

Level 8 signal. Scram trip signals from the turbine retain the

design feature that a single failure will neither initiate nor

impede the initiation of a reactor scram (trip).

c. The setting, function, and conditional requirements of the

APRM Neutron Flux--High, Setdown function are not altered. This

change serves to achieve consistency between two Technical

Specifications Tables. This eliminates the need for surveilling a

function in a mode which is not applicable. The functions of the

APRMs required to be OPERABLE in Operational Condition 1 remain to

ensure that reactor core thermal margins are not compromised.

d. The reference to 4.0.4 applicability will assist to ensure

consistent interpretation of the technical specifications by the

plant operators. This assists in ensuring that the plant is operated

within technical specification limitations. This change does not

affect trip instrumentation setpoints, and the scram function of the

RPS is assured by the weekly functional testing of the Manual Scram.

e. Including the instrumentation channel checks as part of

technical specification requirements provides an appropriately

regimented method of controlling the conduct of the surveillances.

None of the functions, setpoints, or configuration of the associated

analog instrumentation is affected by this administrative change.

f. The administrative nature of the changes serves to provide

more concise guidance to the plant operating staff, and as such do

not impact the safety margin.

The proposed changes do not significantly reduce the margin of

safety as defined in the basis for any Technical Specification.

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: Public Library of Illinois

Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348

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

Austin, One First National Plaza, Chicago, Illinois 60690.

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: January 13, 1995.

Description of amendment request: The proposed amendments would

revise the pressure alarm setpoint allowable values for the emergency

core cooling system (ECCS) and reactor core isolation cooling (RCIC)

system ``keep filled'' pressure instrumentation channels. The purpose

of the proposed change is to lower the setpoint allowable values for

these parameters to more realistic values based upon calculations

performed by the licensee reflecting design changes and system

performance. Also, the term ``setpoint'' is being changed to ``setpoint

allowable value'' to clarify the use of the values. Additionally, two

administrative/editorial changes are included to delete technical

specification footnotes which are no longer applicable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Commonwealth Edison has evaluated the proposed Technical

Specification Amendment and determined that it does not represent a

significant hazards consideration. Based on the criteria for

defining a significant hazards consideration established in 10 CFR

50.92, operation of LaSalle County Station Units 1 and 2 in

accordance with the proposed amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

a. The proposed change in the technical specification allowable

values for the ECCS and RCIC discharge line ``keep filled'' alarm

instrument channels does not change the design bases or function of

these systems as described in the technical specifications and

UFSAR. An analysis performed by engineering demonstrates that the

proposed allowable values are sufficient for verifying that the ECCS

and RCIC pump discharge lines are full of water. In addition,

setpoint [[Page 11130]] calculations have been performed to verify

that sufficient margin exists between the recommended calibration

setpoints and the analytical limits for these instrument channels to

account for all applicable instrument errors. This provides high

assurance that the trip setpoints of these instrument channels will

not drop below the minimum required value. The ``keep filled''

instrumentation is not a factor in the assumptions of any accidents,

thus, the probability of analyzed accidents is not increased.

b. The proposed technical specification amendment does not

revise the configuration of the ECCS and RCIC discharge line ``keep

filled'' instrument channels or sensing lines. The proposed setpoint

allowable values and associated calibration setpoints are within the

calibration ranges of the existing pressure switches. Thus,

implementation of the proposed amendment does not involve any

physical alterations to the plant except for the recalibration of

the pressure switches to the new calibration setpoints.

c. The ECCS and RCIC discharge line ``keep filled'' instrument

channels only perform a monitoring function. Other than ensuring

system readiness they do not perform a function important to safety.

Thus, the probability of a ECCS or RCIC failure is not increased

since the operation and function of the ECCS and RCIC discharge line

fill systems is not affected by this change.

d. The failure of a ECCS or RCIC discharge line fill system will

not go undetected by the proposed change, since water leg pump trips

are annunciated in the control room. In addition, quarterly

surveillances are performed on these pumps to check for degradation.

e. The ECCS and RCIC discharge line fill systems are not used to

mitigate the consequences of an accident or transient. These systems

are not required after the ECCS and RCIC pumps are activated.

Therefore, the proposed change does not cause an increase in the

probability or consequences of an accident previously evaluated.

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

accident from any accident previously evaluated because: This

technical specification amendment only lowers the trip setpoint

allowable values for the ECCS and RCIC discharge line ``keep

filled'' alarm instrumentation channels. As described above, the

proposed setpoint allowable values are sufficient for verifying that

the ECCS and RCIC discharge lines are full of water. Thus, the

probability of a water hammer occurring during system activation for

a surveillance test is not increased. In addition, each instrument

channel is independent from the other channels so that a failure in

one channel will not propagate to another channel. Therefore, the

operation of the facility in accordance with the proposed amendment

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

accident.

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

because: The margin of safety is not affected by this amendment,

because this change involves monitoring instrumentation only. The

purpose of the ECCS and RCIC discharge line ``keep filled'' alarms

is to alert the operators when a ECCS or RCIC system may not be

operable due to empty or partially empty discharge lines. The

proposed amendment does not alter or degrade this function, since

the new setpoint allowable values are adequate for verifying that

the discharge lines are full of water. Therefore the operation of

the facility in accordance with the proposed amendment 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Public Library of Illinois

Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60690.

NRC Project Director: Robert A. Capra.

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

County, Michigan

Date of amendment request: January 13, 1995

Description of amendment request: The proposed amendment would

modify the required settings, and allowable ``as found'' and ``as

left'' tolerances for the primary and secondary safety valves. The

proposed limits would allow installed primary and secondary valve

settings to be within a 3% tolerance of their nominal settings, but

would require returning the valve settings to within 1% of the nominal

settings if the valves are removed from the piping for maintenance or

testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The following evaluation supports the finding that operation of

the facility in accordance with the proposed technical specification

change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to the Technical Specifications increases

the acceptable as found tolerance for the pressurizer safety valves.

The most limiting overpressure event, loss of external load, has

been analyzed to account for this change. The loss of external load

analysis was performed using a conservative 25% steam generator tube

plugging and an initial pressurizer level of 67.8% (providing an

approximate 10% conservative margin above programmed pressurizer

level for full power). Primary and secondary safety valve

accumulation was conservatively accounted for and the setpoint

tolerance of +3% was assumed. Reactor trip on turbine trip was

assumed to be disabled and the atmospheric dump valves were assumed

unavailable. The results of the analysis demonstrated primary and

secondary system pressures within 110% of design pressures.

Therefore, the consequences of overpressurization events will not be

significantly increased with a +3% tolerance on the primary safety

valve setpoints. The proposed Technical Specifications change will

not affect normal plant operation and will not increase the

probability of an accident.

A review of all DNB [departure from nucleate boiling] analyses

was performed to ensure that predicted pressurizer pressures for

those analyses would not be affected by a -3% tolerance on the

lowest setpoint valve. The DNB analyses for which significant

primary system pressure increases were predicted do not result in

pressures high enough to lift the pressurizer safety valves with the

proposed tolerance. A conservative DNB analysis that bounds the

consequences of inadvertent opening of a pressurizer safety valve

has also been previously performed with predicted acceptable

results. If a pressurizer safety valve were to stick open, the

consequences would be bounded by the small break LOCA [loss-of-

coolant accident] analysis. Therefore, the consequences due to a -3%

tolerance on the primary safety valve setpoints will not increase

the consequences or probability of an accident.

The proposed revision removes the requirement for one operable

pressurizer safety valve to be installed whenever the reactor head

is on the vessel. Instead, proposed Specification 3.1.7.1 requires

all pressurizer safety valves to be operable above cold shutdown,

and overpressure protection during cold shutdown is provided by

existing Specification 3.1.8.2, Power Operated Relief Valves.

The proposed Technical Specifications change also lists the lift

settings for each of the primary and secondary system safety valves.

This change will not affect the operation or function of the valves.

Therefore, the probability and consequences of previously evaluated

accidents will not be increased.

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

from any previously evaluated.

The proposed changes to Technical Specifications will not affect

the manner in which the plant operates. The proposed increase in

pressurizer safety valve lift setting tolerance could change the

pressure at which the valves open in an overpressurization event,

but would not create the possibility of a new or different kind of

accident. Since Technical Specification 3.1.8 addresses primary

system overpressurization during cold shutdown, the proposed removal

of the requirement for an operable pressurizer safety valve to be

installed whenever the reactor head is on the vessel will not create

[[Page 11131]] the possibility of a new overpressurization event

during cold shutdown. The proposed change to list the lift settings

for the individual primary and secondary safety valves will have no

effect on the safety function of the valves. Therefore, the proposed

changes will not create the possibility of a new or different kind

of accident from any previously evaluated.

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

The proposed changes to Technical Specifications do not affect

the DNB analyses that have been previously performed. The most

limiting overpressurization event, loss of external load, has been

conservatively analyzed accounting for the proposed changes and

demonstrated that the primary and secondary system pressures remain

within 110% of the design pressures. Overpressurization during cold

shutdown is addressed by Technical Specification 3.1.8. 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: John N. Hannon.

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

County, Michigan.

Date of amendment request: February 10, 1995.

Description of amendment request: The proposed amendment would

modify the Technical Specifications to allow a one time deferral of

several 18-month interval surveillance tests until the upcoming

scheduled refueling outage to avoid the necessity of imposing a plant

shutdown solely for the sake of their performance.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The following evaluation supports the finding that operation of

the facility in accordance with the proposed Technical

Specifications (TS) would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Deferring surveillance testing will introduce no new operating

conditions, change no equipment operating procedures, and change no

plant systems or equipment. Therefore, operation of the facility in

accordance with the proposed TS would not result in a significant

increase in the probability of an accident previously evaluated.

Deferring surveillance testing of snubbers and instrument

channels could allow minor degradations of snubber condition or

small changes in instrument setpoints or calibration to progress

some amount beyond that point which would occur with a shorter

surveillance interval. A review of the recent test history for the

subject surveillance indicates that no significant snubber

degradation or instrument drift was found. It is not expected that,

even with the proposed surveillance deferral, snubber conditions or

instrument settings will be found to exceed conditions allowable by

the Technical Specifications. Therefore, operation of the facility

in accordance with the proposed TS would not result in a significant

increase in the consequences of an accident previously evaluated.

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

from any previously evaluated.

Deferring surveillance testing will introduce no new operating

conditions, change no equipment operating procedures, and change no

plant systems or equipment. Therefore, operation of the facility in

accordance with the proposed TS would not create the possibility of

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

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

A review of past performance of the subject surveillance tests

indicate that the requested deferral of testing would not have a

significant effect on the results of the tests when they are

performed prior to the startup for cycle 12. Most of the affected

instrumentation is monitored each shift by channel checks, which

would disclose major failures or significant drift. Therefore,

operation of the facility in accordance with the proposed TS would

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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: November 2, 1994.

Description of amendment request: The proposed amendment would

delete the content of the Appendix B, Environmental Protection Plan

(EPP) and modify License Conditions 2.C.(2) to delete that portion

which refers to the EPP. Specifically, the requirements for non-

radiological environmental monitoring have been completed. The

radiological environmental monitoring requirements have been

incorporated into Appendix A (the Technical Specifications). There

would be no impact on the continued safety of the McGuire station by

deleting Appendix B.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Deletion of the Environmental Protection Plan and modifying

License Condition 2.C.(2) will have no impact on the probability or

consequences of an accident previously evaluated because the changes

will not have any impact upon the design or operation of any plant

systems or components.

The proposed revision will not create the possibility of a new

or different kind of accident from any previously evaluated because

the revision is administrative in nature and will not change the

types and amounts of effluent that will be released.

The proposed revision will not reduce a margin of safety because

it is administrative in nature and will not effect the margin of

safety as defined in the basis for any Technical Specifications.

Accordingly, this proposed changes does not involve a

significant hazard.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow. [[Page 11132]]

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

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

Date of amendment request: January 18, 1995.

Description of amendment request: The proposed amendments would

relocate the requirements for the seismic instrumentation,

meteorological instrumentation, and loose-part detection system from

the Technical Specifications to the Selected Licensee Commitment (SCL)

Manual. This will allow future changes to these controls to be

performed under the provisions of 10 CFR 50.59. No changes are being

made to the technical content of the affected Technical Specification

pages.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Relocation of the affected TS sections to the SLC Manual

will have no effect on the probability of any accident occurring. In

addition, the consequences of an accident will not be impacted since

the above instrumentation will continue to be utilized in the same

manner as before. No impact on the plant response to accidents will

be created.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No new accident causal mechanisms will be created as a

result of relocating the affected TS requirements to the SLC Manual.

Plant operation will not be affected by the proposed amendments and

no new failure modes will be created.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. No impact upon any plant safety

margins will be created. Relocation of the affected TS requirements

to the SLC Manual is consistent with the content of the Westinghouse

RSTS [Revised Standard Technical Specifications], as the NRC did not

require technical specification controls for the affected

instrumentation in the RSTS. The proposed amendments are consistent

with the NRC philosophy of encouraging utilities to propose

amendments that are consistent with the content of the RSTS.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

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

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

Date of amendment request: January 18, 1995.

Description of amendment request: The amendments would revise

Technical Specification Table 4.3-3 to allow the analog channel

operational test interval for radiation monitoring instrumentation to

be increased from monthly to quarterly. The proposed amendment changes

would be consistent with the guidance in Generic Letter 93-05, ``Line-

Item Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Decreasing the frequency of the radiation monitor analog

channel operational test from monthly to quarterly will have no

impact upon the probability or any accident, since the radiation

monitors are not accident initiating equipment. Analysis of the

previous test data * * * shows that no significant degradation of

performance is to be expected by the decrease in frequency.

Therefore, the requested amendments will have no adverse impact upon

the consequences of any accident.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated above, the radiation monitors are not accident

initiating equipment. No new failure modes can be created from an

accident standpoint. The plant will not be operated in a different

manner.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. Plant safety margins will be

unaffected by the proposed changes. No safety equipment which is

taken credit for in accident analyses will be affected by the

requested amendments. The availability of the affected radiation

monitors will be increased as a result of the proposed amendments

because the monitors will not have to be made unavailable for

testing as frequently. In addition, radiation monitor operating

experience supports the proposed amendments. Finally, the proposed

amendments are consistent with the NRC position and guidance set

forth in NUREG-1366 and Generic Letter 93-05.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: January 20, 1995.

Description of amendment request: The proposed amendments will

relocate the operability requirements for the INCORE DETECTORS (TS 3/

4.3.3.2) to the Updated Final Safety Analysis Report, and revise Linear

Heat Rate surveillance 4.2.1.4, and Special Test Exceptions

surveillances 4.10.2.2, 4.10.4.2 (Unit 2 only), and 4.10.5.2,

accordingly.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes are administrative in nature in that the

specifications for [[Page 11133]] operation and surveillance of the

Incore Instrumentation (ICI) System will be relocated from the

Technical Specifications to the Updated Final Safety Analysis Report

for St. Lucie Unit 1 and Unit 2. Changes to the system will be

controlled by 10 CFR 50.59, and the safety analysis report is

required to be updated pursuant to 10 CFR 50.71(e). Relocation of

these requirements to the UFSAR is consistent with the NRC ``Final

Policy Statement on Technical Specifications Improvements for

Nuclear Power Reactors'' published in the Federal Register (58 FR

39132) dated July 22, 1993.

Incore instrumentation is not an accident initiator nor a part

of the success path(s) which function to mitigate accidents

evaluated in the plant safety analyses. The proposed technical

specification change does not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident, nor do the changes

alter any assumptions or conditions in any of the plant accident

analyses. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment to relocate the existing Technical

Specification requirements for the Incore Instrumentation System to

the Updated Final Safety Analysis Report will not change the

physical plant or the modes of plant operation defined in the

Facility License. The change does not involve the addition or

modification of equipment nor does it alter the design or operation

of plant systems. Therefore, operation of the facility in accordance

with the proposed amendment would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed changes are administrative in nature in that

operating and surveillance requirements for the Incore

Instrumentation System will be relocated from the Technical

Specifications to the Updated Final Safety Analysis Report for St.

Lucie Unit 1 and Unit 2. The ICI system is not used to actuate

safety-related equipment, provide interlocks, or otherwise perform

automatic plant control functions. The system is used to monitor

core power distribution parameters whose limits do involve a margin

of safety; however, the ICI system itself makes no contribution to

that margin of safety, and the power distribution limits will not be

changed by the proposed amendment. Therefore, operation of the

facility in accordance with the proposed amendment would not involve

a significant reduction in a margin of safety.

Based on the above discussion and the supporting Evaluation of

Technical Specification changes, FPL has determined that the

proposed license amendment involves no significant hazards

consideration.

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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: David B. Matthews.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: January 17, 1995.

Description of amendment request: The licensee proposes to revise

the technical specifications to reference Topical Report NF-TR-95-01 as

the documentation of the licensee's proficiency in performing certain

reload design calculations once the NRC has evaluated and approved NR-

TR-95-01.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The addition of the reference to FPL [Florida Power and Light

Company] topical report which demonstrates FPL's ability to perform

certain reload design calculations for Turkey Point Units 3 and 4 is

administrative in nature and has no impact on the probability or

consequences of any Design Bases Event (DBE) occurrences previously

evaluated. The reload design calculations will be performed using

methodologies and computer codes approved by the NRC and poses no

increase in the probability or consequences of any accident

previously evaluated.

The Core Operating Limits Report (COLR) parameters will be

evaluated every cycle to ensure proper compliance with the Updated

Final Safety Analysis Report (UFSAR). These limits will be evaluated

in accordance with 10 CFR [Section] 50.59, which ensures that the

reload will not involve an increase in the probability of

occurrences or consequences of an accident previously evaluated.

Title 10 CFR [Section] 50.59 (2) states that a proposed change

involves an unreviewed safety question (i) if the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety previously evaluated in the safety

analysis report may be increased. Consequently, since any change to

the reload core design analysis must be evaluated relative to the

more restrictive evaluation criterion of 10 CFR [Section] 50.59,

then operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The addition of the reference to FPL topical report which

demonstrates FPL's ability to perform certain reload design

calculations for Turkey Point Units 3 and 4 is administrative in

nature and has no impact, nor does it contribute in any way to the

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

previously evaluated. No new accident scenarios, failure mechanisms

or limiting single failure events are introduced as a result of the

proposed change.

The generation of the Axial Flux Difference, Rod Bank Insertion

limits and K(Z) curve will be performed using NRC-approved

methodology and are submitted to the NRC, as a revision to the COLR,

to allow the NRC staff to trend. The Technical Specifications will

continue to require operation within the core operating limits and

appropriate actions will be taken if these limits are exceeded.

Title 10 CFR [Section] 50.59 permits a licensee to make changes

in the facility as described in the safety analysis report without

prior Commission approval, provided that the proposed changes does

not involve an unreviewed safety question. 10 CFR [Section] 50.59

(2) states that a proposed change involves an unreviewed safety

question (ii) if a possibility for an accident or malfunction of a

different type than any evaluated previously in the safety analysis

report may be created. Consequently, since any change to the reload

core design analysis must be evaluated relative to the more

restrictive evaluation criterion of 10 CFR [Section] 50.59, then

operation of the facility in accordance with the proposed amendments

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

accident from any accident previously evaluated.

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

amendments would not involve a significant reduction in a margin of

safety.

The margin of safety is not affected by FPL performing the

reload design calculations for Turkey Point Units 3 and 4. The

supporting Technical Specification values are defined by the

accident analyses which are performed to conservatively bound the

operating conditions defined by the Technical Specifications. The

development of the limits for future reloads will continue to

conform to the methodology described in NRC approved documentation.

In addition, each future reload will involve a 10 CFR [Section]

50.59 [[Page 11134]] review to assure that operation of the units

within the cycle specific limits will not involve a reduction in a

margin of safety. 10 CFR [Section] 50.59 (2) states that a proposed

change involves an unreviewed safety question (iii) if the margin of

safety as defined in the basis for any technical specification is

reduced. Consequently, since any change to the reload core design

analysis must be evaluated relative to the more restrictive

evaluation criterion of 10 CFR [Section] 50.59, then operation of

the facility in accordance with the proposed amendments would 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. The

NRC staff, however, considers that the licensee's statements relative

to 10 CFR Section 50.59 evaluations to be performed in the future are

not relevant to the proposed no significant hazards determination.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: David B. Matthews.

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

County, Iowa

Date of amendment request: October 28, 1994.

Description of amendment request: The proposed amendment revises

the Duane Arnold Energy Center (DAEC) Operating License by deleting a

condition of the license that requires a ``Plan for Integrating

Scheduling of Plant Modifications for the Duane Arnold Energy Center''

(the Plan).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is provided below:

(1) The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. No physical changes to the facility will occur

as a result of this amendment. Work activities will continue to

receive the appropriate level of review in accordance with DAEC

procedures and practices. The organizational structure that controls

and manages these activities remains unchanged and will assure that

activities are prioritized and performed in a manner consistent with

plant safety. The proposed amendment removes an administrative

burden that is no longer required.

(2) The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No changes to the physical design and/or operation of the

plant will occur as a result of this amendment. The processes by

which activities are planned, prioritized, and controlled are not

affected. The appropriate level of technical review and management

oversight continue to be performed in accordance with existing

procedures and practices to assure that activities are performed in

a manner consistent with plant safety.

(3) The proposed amendment does not involve a significant

reduction in a margin of safety. As stated earlier, no changes to

the physical design and/or operation of any plant systems will occur

as a result of this amendment. Work activities will continue to

receive the appropriate technical review and management oversight to

assure that activities are prioritized and performed in a manner

consistent with plant safety. The amendment removes an

administrative burden that is no longer required.

Based on the above, we have determined that the proposed

amendment will not involve a significant hazards consideration.

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: Cedar Rapids Public Library,

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

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

& Bouckins, 1800 M Street NW., Washington, DC 20036.

NRC Project Director: Leif J. Norrholm.

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

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

Date of amendment request: January 24, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification 3.4.1, ``Leakage Rate,'' to reduce the

allowable leakage rate of the reactor building from 2000 cubic feet per

minute (cfm) to 1600 cfm.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Secondary containment and RBEVS [Reactor Building Emergency

Ventilation System] are not initiators or precursors to an accident.

Secondary containment provides a pressure boundary, with limited in-

leakage, for the purpose of preventing a ground level unfiltered

release of radioactivity. RBEVS responds to accidents involving

release of radioactivity to the secondary containment by maintaining

a negative pressure inside secondary containment and by providing an

elevated release. Therefore, a change to the Reactor Building

leakage rate cannot affect the probability of an accident previously

evaluated.

Although the proposed change reduces the Reactor Building

leakage rate from 2000 cfm to 1600 cfm consistent with system

design, there is no effect on the radiological consequences of any

previously analyzed accident since the radiological analysis does

not assume exfiltration. Therefore, the Technical Specification

change does not significantly increase the consequences of a

previously evaluated accident.

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

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

different kind of accident from any accident previously evaluated.

The proposed change to the Reactor Building leakage rate from

2000 cfm to 1600 cfm does not involve any accident precursors or

initiators. During an accident involving a release of radioactivity

to the secondary containment, the RBEVS would be operable and

provide filtration of containment atmosphere prior to release to the

environment. This change does not involve any physical modifications

to the system, thus the system will operate as designed. Therefore,

the proposed Technical Specification change will not create the

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

previously evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed change in Reactor Building in-leakage from 2000 cfm

to 1600 cfm in Specification 3.4.1 and the associated basis is to be

consistent with system design and reflect the leakage rate

associated with approximately one building air volume change per

day. The resulting accident analysis remains unchanged since the

radiological analysis does not assume any exfiltration. Therefore,

the proposed change will not involve a significant reduction in the

margin of safety as defined in the basis for any Technical

Specification.

Therefore, as determined by the above analysis, this proposed

amendment involves no significant hazards consideration.

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 [[Page 11135]] involves no significant hazards

consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Ledyard B. Marsh.

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

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

Date of amendment request: February 1, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.6.13, ``Remote Shutdown Panels.''

TS 3.6.13 currently requires that if the valve controls or monitoring

instrumentation on the Remote Shutdown Panels are inoperable, they must

be restored to an operable status within 24 hours or the plant shall be

shut down. The proposed change would require inoperable valve control

functions be restored to an operable status within 30 days or the plant

shall be shut down. The proposed change would also specify that

required inoperable monitoring instrumentation functions be restored to

an operable status within 30 days or that an alternate method of

monitoring the parameter be established within 30 days and the required

function be restored to an operable status within 90 days or the plant

shall be shut down.

The proposed amendment would also make minor editorial changes to

TS Table 3.6.13-1 so that the table entries would be consistent with

the proposed revisions to TS 3.6.13.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The remote shutdown panel monitoring instruments and controls

are not initiators or precursors to an accident. The remote shutdown

panels provide the operator with sufficient monitoring instruments

and controls to place and maintain the plant in a safe shutdown

condition from a location other than the control room. Therefore,

the proposed changes to Specification 3.6.13, ``Remote Shutdown

Panels,'' cannot affect the probability of a previously evaluated

accident.

The proposed changes, in part, require that one channel (on

either panel) for each function be operable. This change could

potentially avoid an unnecessary plant shutdown without affecting an

operator's ability to cope with a control room evacuation. One

channel of each function is adequate to assure a safe shutdown. The

proposed changes would also allow 30 days to restore an inoperable

function to an operable status. As indicated in the ITS [Improved

Standard Technical Specifications], the allowed time of 30 days is

acceptable based on operating experience and the low probability of

an event that would require evacuation of the control room. With one

or more monitoring instrument functions inoperable, the proposed

change gives an operator an additional option. Specifically, the

operator is allowed 30 days to establish an alternate method of

monitoring the parameter and 90 days to restore the function to

operable status. The use of an alternate method is acceptable since

it will provide the operator with indication of the parameter of

interest. The remote shutdown panels will not be required to be

operable in hot shutdown because the plant is already subcritical

and in a condition of reduced reactor coolant inventory energy.

Because this Specification no longer applies to hot shutdown and to

be consistent with the guidance provided in the ITS, Specification

3.6.13.d will require that the plant be brought to a hot shutdown

condition (versus cold shutdown condition) in 12 hours. As indicated

in the ITS, the 12-hour completion time is reasonable based on

operating experience. The Bases Section to 3.6.13 and 4.6.13 was

revised to be consistent with the proposed changes to the

Specification. The Bases currently indicates that one remote

shutdown panel is required to be operable. As explained above, one

channel of each required function is required to maintain remote

shutdown operability. In summary, the proposed changes will not

affect the ability of the Remote Shutdown System to provide the

operator with sufficient instrumentation and controls to place and

maintain the plant in a safe shutdown condition from a location

other than the control room. Therefore, the consequences of an event

requiring a control room evacuation will not significantly increase.

Editorial changes were made to Table 3.6.13-1 to be consistent

with the changes made to the Specification. Specifically, the word

``INSTRUMENT'' was changed to ``FUNCTION'' and the words ``PANEL

MONITORING'' were changed to the words ``PANELS FUNCTIONS.'' These

changes make it clear that one channel of each function, on either

panel is acceptable to maintain operability. The emergency condenser

condensate return valve control and motor-operated steam supply

valves control were relocated from Specification 3.6.13.b to Table

3.6.13-1 to be consistent with the proposed changes.

Based on the above, the consequences of an accident previously

evaluated are not significantly increased.

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

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

different kind of accident from any accident previously evaluated.

The changes do not introduce any new accident precursors and do

not involve any alterations to plant configurations which could

initiate a new or different kind of accident. The proposed changes

require that one channel of each function be operable to assure the

remote shutdown panels can meet their intended function. No changes

have been made which will affect the operation of the remote

shutdown panels in a way which would create a new or different kind

of accident. Therefore, the proposed changes will not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed changes will not affect the ability of the Remote

Shutdown System to provide the operator with sufficient

instrumentation and controls to place and maintain the plant in a

safe shutdown condition from a location other than the control room.

The ability to respond to a control room evacuation is maintained

with one channel operable for each required function. The allowed

outage time of 30 days is acceptable based on operating experience

and the low probability of an event requiring control room

evacuation. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Ledyard B. Marsh.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: January 10, 1995.

Description of amendment request: The proposed amendment request

would revise Technical Specifications by deleting the power range,

neutron flux, high negative rate trip from Tables 2.2-1, 3.3-1, and

4.3-1, and delete the associated Bases Section 2.0. [[Page 11136]]

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

* * * The proposed changes would not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The deletion of the power range, neutron flux, high negative

rate trip will not adversely affect plant operations. As has been

presented and accepted by the NRC Staff in previous docketed

correspondence, the dropped RCCA [rod cluster control assembly]

accident analysis does not rely on this trip to safely shut down the

plant. The safety analysis of the plant is unaffected by the

proposed changes. Since the safety analysis is unaffected, the

calculated radiologicalreleases associated with the analysis are not

affected. Therefore, the proposed changes will not increase the

probability or consequences of an accident previously evaluated.

The reactor trip system is used to mitigate accidents. There

have been instances, during calibration of these units, where a

single channel has generated a trip signal. Leaving this in place

when it is not necessary could, therefore, cause a reactor trip. The

deletion of one trip function will, therefore, slightly decease, not

increase, this probability.

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

from any accident previously evaluated.

The reactor trip system is used to mitigate accidents, and the

only way that it can initiate an event is by causing the reactor to

trip when it is unnecessary. This possibility of the generation of a

false trip signal has already been evaluated in the safety analysis.

This modification will physically remove or disable the power range,

neutron flux trip and will therefore decrease the possibility for

the generation of a false trip signal. Therefore, the proposed

change cannot create a new or different kind of accident from any

previously evaluated.

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

The proposed change which deletes the power range, neutron flux,

high negative rate trip will have no impact on the margin of safety.

The current safety analysis for Millstone Unit No. 3 does not credit

this trip for any events; therefore, removal of this trip from the

technical specifications will not affect the margin of safety for

any analyzed events.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: January 23, 1995.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) by 1) adding a new Section 3/

4.5.5 which provides a limiting condition for operation, an action

statement, a surveillance requirement, and a corresponding bases

section, for the trisodium phosphate (TSP) baskets which will be

installed in the next refueling outage; 2) deleting Section 3/4.6.2.3

and Bases 3/4.6.2.3 related to the spray additive system which are no

longer needed since the chemical addition tank is being abandoned; and

3) updating Index Pages viii, ix, and xiv to reflect the above changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

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

changes would not:

1. Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

The plant change affects the chemical composition of the QSS

[quench spray system] flow and the method of sump pH control, which

are important for containment heat removal/pressure mitigation (MSLB

and LOCA) [main steamline break and loss-of coolant accident] and

fission product removal (LOCA). However, this change does not affect

the probability of occurrence of these accidents. Since the TSP

baskets are passive devices located inside the containment, they

cannot initiate a transient or affect the probability of occurrence

of any previously evaluated accident.

The design change will not adversely affect the radiological

doses for the DBA [design basis accident] LOCA at the Exclusion Area

Boundary, Low Population Zone, Millstone Unit No. 3 Control Room,

Millstone Unit No. 2 Control Room, and the Millstone Technical

Support Center. Also, the change will not adversely affect the

calculated peak clad temperature (PCT) for the DBA LOCA.

2. Create the Possibility of a New or Different Kind of Accident

from any Previously Analyzed.

The change does not create a malfunction that is different from

those previously evaluated. The TSP baskets are passive devices that

have minimal impact on any other systems except through water

chemistry. The change in water chemistry does not adversely affect

any safety systems. The installation of the TSP baskets and the

abandonment of the CAT [chemical addition tank] will not change the

probability of a malfunction of safety-related equipment.

Potential malfunctions relating to the TSP powder, the 12

baskets which hold the TSP powder, the QSS and other systems, and

equipment credited in the safety analysis were evaluated and

determined not to be adversely affected by the change. Additionally,

the transient pH behavior of the spray flow will not adversely

affect metals, coatings and elastomers in the containment, and the

performance of associated safety functions is not affected.

Finally, the change in the chemical composition of the QSS

solution will not affect the operability of this system or its

ability for containment heat removal and pressure mitigation.

3. Involve a Significant Reduction in the Margin of Safety.

The design changes do not adversely affect the ability of the

QSS to perform the function of containment heat removal, pressure

mitigation and fission product (iodine) retention. The design

changes do not adversely affect any equipment credited in the safety

analysis. Also, the design changes to not increase the calculated

peak clad temperature (PCT) or the offsite doses due to the design

basis LOCA. Therefore, there is no impact on the margin of safety as

specified in the technical specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: January 24, 1995. [[Page 11137]]

Description of amendment request: The amendment request would

revise the Technical Specification Section 3.2.3.1.a and Table 2.2-1 to

decrease the acceptance criterion for measured reactor coolant system

(RCS) flow rate from 387,480 gallons per minute (gpm) to 371,920 gpm.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

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

changes would not:

1. Involve a Significant Increase in the Probability or

Consequence of an Accident Previously Evaluated.

An evaluation of the 4% decrease in the RCS total flow rate

limit has shown that the change does not significantly impact the

design basis analyses. Therefore, the change will not increase the

consequences of an accident previously evaluated.

There are no actual plant changes that will result from this

technical specification change. Instead, the technical specification

requirement for minimum total RCS flow rate is being changed to

provide operational benefit without compromising safety. Since there

are no plant changes, there is no effect on the probability of

occurrence of previously evaluated accidents.

The change will have a negligible impact on the small break loss

of coolant accident (LOCA) and large break LOCA analyses. The PCT

[peak cladding temperature] acceptance criteria will continue to be

met with the assumption of a 4% reduction in RCS flow rate.

For the steam generator tube rupture event, both the FSAR [Final

Safety Analysis Report] offsite dose analysis and the margin of

steam generator (SG) overfill were evaluated. It was determined that

the 4% reduction in RCS flow rate will not adversely affect the

offsite doses or the margin to SG overfill and, therefore, the FSAR

conclusions remain unchanged.

In the evaluation of non-LOCA transients, the DNB [departure

from nucleate boiling] is the most affected parameter due to a

change in flow rate. It was concluded that the 4% reduction in RCS

flow was acceptable and there was margin to the DNB limit.

It is concluded that there is sufficient margin to the system

pressure, PCT and DNB limits to offset the effect of the 4% flow

rate decrease and the calculated radiological releases associated

with the analysis are not affected. Therefore, there is no effect on

the consequences of previously evaluated accidents.

2. Create the Possibility of a New or Different Kind of Accident

from any Previously Analyzed.

The low loop flow trip setpoint specified in Technical

Specification Table 2.2-1 is set as a fraction of total flow. The

flow fraction is not being changed and no hardware changes are

required due to the reduction in minimum flow. Also, the reduction

in minimum flow will not change the operation of any plant equipment

and it does not modify plant operation.

Therefore, the reduction in minimum flow does not introduce any

new failure modes or malfunctions and it does not create the

potential for a new unanalyzed accident.

3. Involve a Significant Reduction in the Margin of Safety.

The proposed 4% decrease in the technical specification limit

for total RCS flow rate will not adversely affect the results of the

FSAR accident analysis, and it is concluded that this change is

safe. The change does not adversely affect any equipment credited in

the safety analysis, and it does not affect the probability of

occurrence of any plant accident. Also, the change has a negligible

impact on the PCT, and it does not increase the offsite doses or

decrease the DNB below its acceptance limit.

Therefore, the change does not have any significant impact on

the protective boundaries, and there is no reduction in the margin

of safety as specified in the technical specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

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

Unit No. 1, Washington County, Nebraska

Date of amendment request: January 9, 1995.

Description of amendment request: The proposed amendment to the

technical specifications (TSs) would delete requirements for the toxic

gas monitoring system (TGMS) as contained in TS 2.22 and TS 3.1, Table

3-3, item 29.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

The previously evaluated accidents affected by this change are

the on-site and off-site toxic chemical releases. These events have

been re-evaluated for this proposed change and have been shown to

meet the applicable regulatory screening criteria. The deterministic

analyses performed show that the guidelines of Regulatory Guide 1.78

for control room habitability are met for on-site and most off-site

chemicals. On-site chemical sources originally present when the

toxic gas monitoring system was installed have been removed from

site or determined not to exceed the deterministic analysis

screening requirements. For those off-site chemical releases which

did not meet the deterministic screening criteria a probabilistic

analysis was performed. The probabilistic analysis performed in

support of this proposed change shows that the probability of an

off-site chemical release leading to 10 CFR 100 consequences is

orders of magnitude less than the SRP [Standard Review Plan] 2.2.3

guidelines. These results show that there is no significant increase

in the probability or consequences of any accident previously

evaluated.

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

or different kind of accident from any previously analyzed.

Only events involving chemicals for which the TGMS provides an

automatic detection/isolation function are affected by this change.

As stated above, the potential events involving these chemicals have

been re-evaluated using the appropriate regulatory guidance and

shown to satisfy either the deterministic screening criteria of RG

[Regulatory Guide] 1.78, or to be probabilistically insignificant

compared to the guidelines of SRP Section 2.2.3. These results show

that the proposed change will not create the possibility of a new or

different kind of accident from any previously evaluated. Therefore,

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

different kind of accident from any previously analyzed.

(3) The proposed changes do not involve a significant reduction

in a margin of safety.

The margin of safety is defined by the regulatory basis for the

existing TGMS, namely NUREG-0737, Item III.D.3.4. The analysis

provided to support this proposed change follows the regulatory

guidelines of RG 1.78 and SRP Section 2.2.3, as specified in NUREG-

0737, Item III.D.3.4. The analysis shows that the applicable

regulatory criteria are met and 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: LeBoeuf, Lamb, Leiby, and MacRae, 1875

Connecticut [[Page 11138]] Avenue, NW., Washington, DC 20009-5728.

NRC Project Director: Theodore R. Quay.

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

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

California

Date of amendment requests: February 6, 1995 (Reference LAR 95-01).

Description of amendment requests: The proposed amendments would

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

Power Plant, Unit Nos. 1 and 2, to change TS 3/4.9.14.1, ``Spent Fuel

Assembly Storage,'' TS 3/4.9.14.2, ``Spent Fuel Pool Boron

Concentration,'' TS 5.3.1, ``Reactor Core--Fuel Assemblies,'' and TS

5.6.1, ``Fuel Storage--Criticality,'' and add new TS 3/4.9.14.3,

``Spent Fuel Assembly Storage--Spent Fuel Pool Region 1.'' The specific

TS changes proposed are as follows:

(1) The proposed changes to TS 3/4.9.14 are:

(a) TS 3.9.14.1 and Figure 3.9-2 would be revised to allow the

storage of spent fuel assemblies with initial enrichments up to 5.0

weight percent uranium-235 (U-235) in Region 2 of the spent fuel pool

(SFP). Fuel pellet diameter would be considered in combination with

initial enrichment and cumulative burnup.

(b) Editorial corrections to the titles of TS 3/4.9.14.1 and 3/

4.9.14.2 would be made for consistency with the TS format.

(2) New TS 3/4.9.14.3 would be added. The new TS would include:

(a) Requirements for acceptable fuel storage in Region 1 of the

SFP.

(b) An action statement, similar to that for TS 3.9.14.1, requiring

suspension of all fuel movement and crane operations except to move the

noncomplying fuel assemblies into an acceptable pattern. The action

statement also requires verification of SFP boron concentration at

least once per 8 hours.

(c) A requirement, similar to that for TS 4.9.14.1, for an

evaluation that considers enrichment, boron content, and cumulative

burnup of each fuel assembly before storage in Region 1 of the SFP.

(d) New Figure 3.9-3 for use in determining the acceptability of

storing fuel in Region 1 of the SFP.

(3) The proposed changes to TS 5.3.1 are:

(a) The number of fuel rods in each fuel assembly, nominal length

of each fuel rod, and maximum fuel enrichment would be removed.

(b) The current allowance for fuel rod substitutions as justified

by analysis would be clarified to specify that the analysis be

performed using NRC staff-approved methods.

(c) An allowance to use a limited number of lead test assemblies in

nonlimiting core locations would be added.

(d) The current specification requiring Zircaloy-4 fuel cladding

would be changed to allow Zircaloy-4 or ZIRLO cladding.

(4) The proposed changes to TS 5.6 are:

(a) TS 5.6.1.1 would be renumbered TS 5.6.1 and the word

``borated'' would be replaced with ``unborated.''

(b) A new requirement would be added to specify the maximum fuel

enrichment allowed to be stored in the fuel racks.

(c) TS 5.6.1.2 would be deleted.

(5) The associated Bases would also be appropriately revised.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Analyses were performed to verify that an increase in enrichment

of the fuel from 4.5 weight percent U-235 to 5.0 weight percent U-

235 would not result in an inadvertent criticality event in the new

fuel storage racks or the SFP. The analyses indicate that for the

new fuel racks, the keff will remain below 0.95 if flooded with

non-borated water, and below 0.98 if flooded with optimum-density

aqueous foam. The analyses indicate that for the spent fuel racks,

assuming credit for soluble boron in accident scenarios, the

keff will remain below 0.95 as required.

The increase in the fuel enrichment from 4.5 weight percent U-

235 to 5.0 weight percent U-235 does not change any of the external

dimensional characteristics of the fuel element, the fuel storage

racks, or the SFP itself. The accidents originally evaluated

considered those events that could lead to fuel damage and release

of radioactive material primarily from mechanical means, such as

physical impact on the fuel or the SFP. Because the physical design

and methods of operation are the same as previously evaluated, there

is no change in the probability of occurrence of such events.

The maximum spent fuel gap activity and the resulting offsite

dose consequences after a postulated fuel handling accident are

primarily dependent on fuel burnup, and are not significantly

affected by an increase in fuel enrichment. For up to 5.0 weight

percent U-235 and 60,000 MWD/MTU burnup, NUREG/CR-5009 indicates

that fuel handling accident offsite doses could increase by a factor

of 1.2, which indicates that doses would still remain within 10 CFR

Part 100 limits.

The Generic Letter 90-02 Supplement 1 change to TS 5.3.1

clarifies the requirements associated with fuel reconstitution. It

does not change the methodology that would be used to reconstitute

fuel.

The use of ZIRLO cladding will not increase the probability or

consequences of an accident, since it has improved mechanical

properties such as a lower corrosion rate and reduced radiation-

induced growth.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The physical and mechanical parameters associated with the fuel

assemblies and spent fuel racks are the same as previously

evaluated. Therefore, any malfunctions related to the physical

aspects of fuel storage are the same as previously evaluated.

The conditions for fuel storage in the proposed new TS 3.9.14.3

provide new criteria for locations where a fuel assembly could be

incorrectly placed. However, the incorrect placement of a fuel

assembly has been analyzed, and would not cause an inadvertent

criticality or any other accident.

The change to 5.0 weight percent U-235 does not result in

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

change the method by which any safety-related system performs its

function. The use of ZIRLO cladding does not result in a significant

change to the plant.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The acceptance criteria of a keff of 0.95 (or 0.98 for the

new fuel rack optimum moderation accident) provides the margin to

criticality. Analyses were performed that conclude that the proposed

changes to allow up to 5.0 weight percent U-235 in the new and spent

fuel racks meet the acceptance criteria. The use of ZIRLO cladding

will not reduce the protection of the public health or safety, as

indicated in the NRC's revisions to 10 CFR 50.44 and 50.46 (57 FR

39355).

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

[[Page 11139]] Department, San Luis Obispo, California 93407.

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

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

NRC Project Director: Theodore R. Quay.

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

Plant, Unit 3, Humboldt County, California

Date of amendment request: November 23, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications Section VI, ``Waste Disposal

Systems,'' regarding radioactive effluent limitations and the

conditions for automatically pumping the contents of the reactor

caisson sump to the outfall canal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed revisions to the HBPP Technical Specifications

remove the ambiguity in the guidelines for directing caisson sump

discharges to the outfall canal. Additionally, the proposed

revisions will modify Section VI to be consistent with the guidance

provided by NRC Draft Generic Letter for 10 CFR 20 Modification to

Technical Specifications (58 FR 68171, dated December 23, 1993).

These changes in effluent limits are not related to the probability

or consequences of an accident.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed revisions to the HBPP Technical Specifications are

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

safety-related system performs its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

The proposed revisions to the HBPP Technical Specifications do

not affect the margin of safety associated with parameters for any

accident analysis.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Humboldt County Library, 636 F

Street, Eureka, California 95501.

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

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

NRC Project Director: Seymour H. Weiss.

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

Plant, Unit 3, Humboldt County, California

Date of amendment request: November 23, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications Section VII.C, Plant Staff, to

decrease the minimum staff requirements for the shift operating

organization from five to two persons.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The probability or consequences of an accident previously

evaluated will not be affected by the change in plant staffing. The

plant staff manning requirements for the shift operating

organization are being reduced to reflect the condition of the plant

in a SAFSTOR mode. Previously evaluated accidents do not require

operator actions to mitigate or reduce the consequences of

occurrence. Consequently, the change will not affect the probability

or consequences of an accident occurring.

Therefore, the proposed 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 revisions to the HBPP Technical Specifications are

administrative in nature. Further, there would not be any change in

equipment or system function or operation.

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 proposed revisions to the HBPP Technical Specifications do

not affect the margin of safety of any accident analysis since they

do not affect the parameters for any accident analysis, and they

have no effect on the current operating methodologies or actions

that govern plant performance.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Humboldt County Library, 636 F

Street, Eureka, California 95501.

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

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

NRC Project Director: Seymour H. Weiss.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket No.

50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County,

Pennsylvania

Date of application for amendment: January 13, 1995.

Description of amendment request: The proposed changes revise

Tables 3.7.1 and 3.7.4 to reflect a reduction in the number of primary

containment power operated outboard valves for the Traversing Incore

Probe (TIP) probes, and a redesignation of the containment penetration

numbers for the TIP ball, shear, and check valves. The proposed changes

are a result of PBAPS Modification P00068.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The TIP system does not serve as an initiator or contributor to

any accidents previously evaluated. The system provides a means of

calibrating the Local Power Range Monitors and supports thermal

limit calculations. The new system performs the same function as the

old one. It will provide improved reliability and added redundancy

by allowing a complete flux mapping if a detector or drive failure

were to occur.

Installation of Modification P00068 and its operation will not

degrade any active or passive equipment that responds to an

accident. These changes do not decrease the

[[Page 11140]] effectiveness of equipment relied upon to mitigate

the previously evaluated accidents.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any previously evaluated.

The modification is considered an enhancement to the TIP system

and does not serve as an initiator or contributor to any of the

accidents previously evaluated. The proposed changes do not

introduce a new mode of plant operation. The new system, like the

old one, is designed to keep the ball valves closed upon reset of

the Primary Containment Isolation System (PCIS) logic. The new TIP

control console will respond to a PCIS isolation signal in the same

manner as the old system.

Implementation of the proposed changes will not affect the

design function or configuration of any component or introduce any

new operating scenarios or failure modes or accident initiation.

Modification P00068 will not impair or prevent safety systems

from performing their safety function. It will not make any changes

to the design function of the TIP system. The classification of the

TIP ball and shear valves and their control circuitry will not

change as a result of this modification.

Therefore, the proposed changes do not create the possibility of

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

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The TIP system does not serve as an initiator or contributor to

any accidents evaluated in the SAR [safety analysis report].

Modification P00068 is considered an enhancement to the existing TIP

system and does not change its design function. The reduction in the

number of containment penetrations from five to three does not

represent a reduction in a margin of safety because of additional

indexers in the new system. The proposed changes do not adversely

affect the assumptions or sequence of events used in any accident

analysis.

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

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: June 13, 1994.

Description of amendment request: The proposed change would remove

license condition 2.E from the Facility Operating License. License

Condition 2.E incorporated the requirements of U.S. Department of

Interior publication ``Environmental Criteria for Electric Transmission

Systems''--1970, which applies to the construction cleanup,

restoration, and maintenance of transmission lines.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

(1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change will remove a license condition unrelated to

nuclear safety. License condition 2.E incorporated into the

Operating License the requirements of U.S. Department of Interior

publication ``Environmental Criteria for Electric Transmission

Systems''--1970. The goal of this standard is to ``safeguard

aesthetic and environmental values within the constraints imposed by

the current state of high-voltage transmission technology.'' License

condition 2.E addresses the preservation of the environment and

natural resources. Removing this condition from the Facility

Operating License has no bearing on plant safety or the health and

safety of the public considering its non-nuclear nature. The

transmission line right-of-ways maintained by the [Power] Authority

[of the State of New York] are subject to regulation by other State

and Federal Agencies. Removal of this license condition will not

affect operation of safety related structures, systems or components

nor affect the quality assurance program at the FitzPatrick plant.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

accident from any accident previously evaluated.

License condition 2.E of the James A. FitzPatrick Plant

Operating License applies to the construction cleanup, restoration,

and maintenance of transmission lines. The Authority's transmission

lines are managed under guidelines based on the ``Generic

Transmission Line Right-of-Way Management'' plan requirements. The

requirements imposed by the plan on the FitzPatrick transmission

line right-of-ways exceed those of the U.S. Department of Interior

publication referenced in license condition 2.E in both scope and

details. Therefore, implementing the proposed change will not create

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

accident previously evaluated.

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

License condition 2.E of the James A. FitzPatrick Plant

Operating License applies to the construction cleanup, restoration,

and maintenance of transmission lines. The requirements imposed by

this license condition are unrelated to nuclear safety. Continued

operation of the plant without Condition 2.E does not involve a

significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Ledyard B. Marsh.

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

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

Date of amendment request: December 16, 1994; supplemented February

10, 1995 (TS 94-07).

Description of amendment request: The proposed change would reduce

the maximum allowed power levels and more clearly specify the plant

conditions allowed by the technical specifications for operation with

one or more main steam safety valves inoperable. In addition, the Bases

would be revised to reflect these changes and incorporate the revised

methodology used to establish the neutron flux setpoints.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of [[Page 11141]] Sequoyah Nuclear Plant (SQN)

in accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

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

inoperable, to ensure that the secondary system is not

overpressurized during the most severe pressurization transient of

the secondary side. Additionally, this change will combine the TS

action statements for 3- and 4-loop operation with one or more MSSVs

inoperable, revise the mode requirements and times of Action

Statement 3.7.1.1.a, and correct a reference in the bases section to

Table 3.7-1. Reduction of the high neutron flux (HNF) trip setpoint

will continue to be used as the means to ensure that the required

reactor power level reductions are met. Mode 3 will be limited to

application when the reactor trip breakers (RTB) are closed. Lack of

NIS trip setpoint adjustments with the RTB open has no effect on the

accident analysis. There is no change to the function of the MSSVs

by the proposed change. This change will not alter any accident

analysis assumptions or results for SQN. The proposed change will

reduce the amount of relief capacity required to mitigate the

consequences of the transient by reducing the total amount of energy

in the primary system. Therefore, this change will not increase the

probability of an accident.

This change is consistent with current SQN accident analysis

assumptions for the MSSVs and does not change the containment

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

mitigation of an accident will result from this proposed change and

no change will occur in the consequences of any accident currently

analyzed.

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

from any previously analyzed.

Inadvertent opening of a MSSV is currently analyzed as an

initiating event for accidental depressurization of the main steam

system. The proposed change does not alter the valves or any other

plant component. The valves will continue to perform as analyzed in

current accident analyses. The proposed change will not create the

possibility for any new or different kind of accident.

By retaining the use of the HNF trip setpoint reduction, no

change is being proposed in the methodology used to ensure that

power reductions are carried out; therefore, this will not create

the possibility of placing the plant into any new unanalyzed

condition. Not adjusting the Nuclear Instrumentation System trip

setpoint with the RTBs open will not create an accident. The

existing accident analysis is still bounding.

Combining the separate action statements for 3- and 4-loop

operation into a single action does not create the possibility for a

new or different kind of accident. Operation with 4 loops will

continue to be required in Modes 1 and 2 by TS 3.4.1.1.

Operation with less than 4 loops will continue to be governed by

TS 3.4.1.2 in Mode 3 and TS 3.4.1.3 in Mode 4. This change will not

place the plant in a configuration not currently bounded by existing

accident analysis.

Revising the mode requirements and their associated times,

consistent with the requirements in NUREG-1431, will continue to

ensure that if the unit is unable to comply with the limiting

condition for operation, the unit will begin an orderly shutdown

until a mode is reached where the specification is not applicable.

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

The proposed change reduces the total energy of the reactor

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

their intended function as assumed in current accident analyses.

This change has been evaluated on a generic basis for Westinghouse

Electric Corporation designed 4-loop nuclear steam supply systems.

SQN plant specific features have been evaluated including power

limit calculations and the interaction of the reactor protection

system trip time delay and the anticipated transient without scram

mitigating system actuation circuitry. Correcting this

nonconservatism restores the margin of safety to what was originally

envisioned. Therefore, the margin of safety assumed in the accident

analysis is not reduced by this change.

Combining the separate action statements for 3- and 4-loop

operation into a single action has no effect on the margin of safety

for 4-loop operation with one or more MSSVs inoperable. Under the

revised TS, 3-loop operation with one or more MSSVs inoperable would

only be allowed in Mode 3, and 4-loop operation will be required in

Modes 1 and 2 in accordance with current TSs 3.4.1.1 and 3.4.1.2.

Revising the mode requirements and their associated times,

consistent with the requirements in NUREG-1431, will not reduce the

safety margin since the new requirements will continue to place the

unit in a mode where the TS is no longer applicable. The new

completion times for mode changes are reasonable, based on operating

experience, to reach the required unit conditions from full power

conditions in an orderly manner without challenging unit systems.

The margin of safety is unaffected by modifying the limits of

Mode 3 applicability to require the RTBs to be closed as the

intended safety function is already completed when they are open.

The NRC 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.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

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

Callaway County, Missouri

Date of amendment request: December 9, 1994, and January 27, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirement 4.6.1.2.a

and its associated Bases. The changes would defer the next scheduled

containment integrated leak rate test (CILRT) for one outage, from

Refuel 7 (March 1995) to Refuel 8 (scheduled for September 1996).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety previously evaluated in the Safety

Analysis Report.

The Callaway CILRT history provides substantial justification

for the proposed test schedule. Three CILRTs have been performed

over a seven year period with successful results. The tests indicate

that Callaway has a low leakage containment. There are no structural

mechanisms which would adversely affect the structural capability of

the containment and that would be a factor in extending the CILRT

schedule by one refueling outage.

A risk impact assessment was performed, and a determination was

made that there is insignificant risk impact as a result of changing

the CILRT schedule. Containment leak rate testing is not an

initiator of any accident, the proposed interval extension does not

affect reactor operations or the accident analysis, and has no

radiological consequences. There will be no changes to 10 CFR 100

dose limits or the control room dose limits. Extending the test

interval will not, by itself, increase the probability of a

malfunction of equipment important to safety. Therefore, the

proposed change will not involve a significant increase in the

probability or consequences of any accident previously evaluated.

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

from any previously evaluated in the Safety Analysis Report.

There are no design changes being made that would create a new

type of accident or malfunction. The proposed change will not alter

the plant or the manner in which it is operated. The change revises

the schedule for performing the periodic CILRT. The purpose of the

test is to provide periodic verification of the leaktight integrity

of the primary reactor containment, and systems and components which

penetrate containment. The tests assure that leakage through

containment and systems and components penetrating containment will

not exceed the allowable leakage rate values associated with

[[Page 11142]] conditions resulting from an accident. The change in

schedule for performing the CILRT will not adversely affect the

containment integrity in the event of an accident. Therefore, the

proposed change will not create the possibility of a new or

different type of accident from any accident previously evaluated.

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

The proposed change to the schedule for performing the periodic

Type A test does not reduce the margin of safety assumed in the

accident analysis for any release of radioactive materials or reduce

any margin of safety preserved by the technical specifications. The

methodology, acceptance criteria, and the technical specification

leakage limits for the performance of the Type A tests will not

change. The Type A tests will continue to be performed in accordance

with 10 CFR 50, Appendix J and the Callaway Technical

Specifications. Therefore, the proposed change will not involve a

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: Leif J. Norrholm.

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: January 24, 1995.

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Section 15.6.5, ``Review and

Audit,'' and TS Section 15.7.8, ``Administrative Controls.'' The

quality assurance audit frequencies would be removed, the section on

emergency plan reviews would be removed, and the period for radioactive

effluent reporting would be increased to annual. In addition, the

references to ``Semiannual Monitoring Report'' would be changed to

``Annual Monitoring Report'' throughout TS Section 15.7.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

In accordance with the requirements of 10 CFR 50.91(a),

Wisconsin Electric Power Company (Licensee) has evaluated the

proposed changes against the standards of 10 CFR 50.92 and has

determined that the operation of Point Beach Nuclear Plant, Units 1

and 2, in accordance with the proposed amendments, does not present

a significant hazards consideration.

A proposed facility operating license amendment does not present

a significant hazards consideration if operation of the facility in

accordance with the proposed amendment will not:

1. Create a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any accident previously evaluated.

3. Will not create a significant reduction in a margin of

safety.

The proposed changes are administrative in nature. There is no

physical change to the facility, its systems, or its operation.

Since the changes will allow more flexibility in assigning resources

to work on poor or weak performance areas, the plant safety will be

enhanced. Operation of PBNP in accordance with the proposed

amendments cannot create an increase in the probability or

consequences of an accident previously evaluated, create a new or

different kind of accident, or result in a significant reduction in

a margin of safety. Therefore, the proposed changes do not present a

significant hazards consideration.

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: Leif J. Norrholm.

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.

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

Unit 1, York County, South Carolina

Date of amendment request: October 18, 1994.

Description of amendment request: The proposed amendment would

change Technical Specification 3.6.1.2 to defer the next scheduled

containment integrated leak rate test at Catawba Unit 1 for one outage,

from the end-of-cycle (EOC) 8 refueling outage (scheduled for February

1995) to EOC 9 (scheduled for June 1996).

Date of publication of individual notice in Federal Register:

February 6, 1995 (60 FR 7073).

Expiration date of individual notice: March 8, 1995.

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

Black Street, Rock Hill, South Carolina.

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

Unit 1, York County, South Carolina

Date of amendment request: November 29, 1994, as supplemented

January 12 and 27, 1995.

Description of amendment request: The proposed amendment requested

renewal for Catawba Unit 1 Cycle 9 operation of the steam generator

tube inspection bobbin probe voltage-based interim plugging criteria

that had been previously approved for Cycle 8.

Date of publication of individual notice in Federal Register:

February 9, 1995 (60 FR 7801).

Expiration date of individual notice: March 13, 1995.

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

Black Street, Rock Hill, South Carolina.

Georgia Power Company, et al., Docket Nos. 50-424 and 50-425, Vogtle

Electric Generating Plant, Units 1 and 2, Burke County, Georgia,

Date of amendment request: January 20, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification 6.4.1.2 to provide a more accurate

description of the Plant Review Board composition.

Date of publication of individual notice in Federal Register:

February 6, 1995 (60 FR 7077). [[Page 11143]]

Expiration date of individual notice: March 8, 1995.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

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: November 7, 1994, as supplemented by

letters dated December 20, 1994, and January 23, 1995.

Brief description of amendment request: The proposed amendments

would change the number of diesel generators (emergency power supply)

required to be operable during Mode 6 with greater than or equal to 23

feet of water above the reactor vessel flange, from two to one. The

amendments would also allow limited substitution of an alternate onsite

emergency power source for one of the two required diesel generators,

in Mode 5 and in Mode 6 with less than 23 feet of water. In addition,

changes to certain system specifications that are affected by the

changes for the emergency power supply were also proposed.

Date of individual notice in Federal Register: January 30, 1995 (60

FR 5739).

Expiration date of individual notice: March 1, 1995.

Local Public Document Room location: Wharton County Junior College,

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

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

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

Date of amendment request: January 27, 1995.

Brief description of amendment request: The amendment modifies the

technical specifications (TSs) by eliminating selected response time

testing as described in the BWROG topical report NEDO-32291, ``System

Analyses for Elimination of Selected Response Time Testing

Requirements.'' The affected TSs are TS 3.3.1.1, ``Reactor Protection

System (RPS) Instrumentation,'' TS 3.3.5.1, ``Emergency Core Cooling

System (ECCS) Instrumentation,'' TS 3.3.6.1, ``Primary Containment and

Drywell Isolation Instrumentation,'' and TS 3.5.1, ``ECCS--Operating.''

Date of publication of individual notice in Federal Register:

February 3, 1995 (60 FR 6739).

Expiration date of individual notice: March 6, 1995.

Local Public Document Room location: Vespasian Warner Public

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

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: October 31, 1994, as

supplemented by letter dated December 28, 1994.

Brief description of amendments: The amendments revise the

refueling machine overload cutoff limit from less than or equal to 1556

pounds to less than or equal to 1600 pounds. The change was requested

because design and fabrication improvements have increased the weight

of the fuel assembly.

Date of issuance: February 9, 1995.

Effective date: February 9, 1995, to be implemented within 45 days

of the date of issuance.

Amendment Nos.: 89, 76, and 60.

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 6, 1995 (60 FR

2160). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 9, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

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

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

and 3, Maricopa County, Arizona

Date of application for amendments: November 20, 1992, as

supplemented by letters dated October 22, 1993, and November 30, 1994.

Brief description of amendments: The amendments would increase the

allowable out-of-service time for the core operating limit supervisory

system (COLSS) from 1 hour to 4 hours before the more restrictive

limits based on the core protection calculators (CPCs) must be applied.

Date of issuance: February 14, 1995.

Effective date: February 14, 1995.

Amendment Nos.: 90, 77, and 61.

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 6, 1993 (58 FR

591) The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 14, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of application for amendment: September 6,

1994. [[Page 11144]]

Brief description of amendment: The amendment would remove

Technical Specification Section 4.5.H.4 which requires the testing and

calibration of pressure switches in certain emergency core cooling

system lines.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995.

Amendment No.: 157.

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

Technical Specifications.

Date of initial notice in Federal Register: October 26, 1994 (59 FR

53838). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 26, 1994 (59 FR 53838).

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: March 25, 1994, as supplemented

on July 29, 1994, and August 24, 1994.

Brief Description of amendments: The amendments change the

Technical Specifications to correct several typographical errors, to

incorporate material implicitly contained in a footnote to an

applicability statement, to provide detailed labels for items listed in

a table, to correct the citation of references, and to remove

references to the Rod Sequence Control System that should have been

included in a previous change.

Date of issuance: February 1, 1995.

Effective date: February 1, 1995.

Amendment Nos.: 174 and 205.

Facility Operating License Nos. DPR-71 and DPR-62. Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: May 25, 1994 (59 FR

27050). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 1, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

Date of application for amendment: December 12, 1994.

Brief description of amendment: The amendment revises the

containment spray (CS) nozzle surveillance interval from 5 to 10 years.

Date of issuance: February 10, 1995.

Effective date: February 10, 1995.

Amendment No.: 157.

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

Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

497).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 10, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Hartsville Memorial Library,

147 West College, Hartsville, South Carolina 29550.

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

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

Date of application for amendments: June 24, 1994.

Brief description of amendments: The amendments revise the

Technical Specifications by deleting the containment recirculation sump

level from Accident Monitoring Instrumentation Tables 3.8.9-1 and

4.8.9-1.

Date of issuance: February 9, 1995.

Effective date: February 9, 1995.

Amendment Nos.: 160 and 148.

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

revised the Technical Specifications.

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

37066).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 9, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

Plant, Middlesex County, Connecticut, and Northeast Nuclear Energy

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

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

Date of application for amendments: June 30, 1994, as supplemented

November 18, 1994, and January 12, 1995.

Brief description of amendments: The amendments modify the

Administrative Controls Section of the Technical Specifications by

replacing the present Nuclear Review Board (NRB) for the Haddam Neck

Plant, and the NRB and site Nuclear Review Board for Millstone Station

with a Nuclear Safety Assessment Board which will serve Millstone Units

1, 2, and 3, and Haddam Neck.

Date of issuance: February 14, 1995.

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

30 days.

Amendment Nos.: 181, 79, 184, 104.

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

Amendments revised the Technical Specifications.

The November 18, 1994, and January 12, 1995, submittals provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45021).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 14, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457, for the Haddam Neck Plant, and Learning

Resource Center, Three Rivers Community-Technical College, Thames

Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for

Millstone 1, 2, and 3.

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

Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: July 29, 1994, as supplemented

in a letter dated December 13, 1994.

Brief description of amendment: This amendment revises Technical

Specifications (TSs) 3/4.4.5 and 3.4.6.2 including associated Bases 3/

4.4.5 and 3/4.4.6.2 to allow the implementation of [[Page 11145]] steam

generator tube interim plugging criteria (IPC) for the tube support

plate elevations during operating cycle 11. The current TSs require

that tubes with imperfections exceeding 40 percent of the nominal tube

wall thickness be removed from service. The IPC will allow a test

voltage-based criterion of 1.0 volts as determined by a bobbin probe

inspection of the tubes. Voltages greater than 1.0 volt will be further

examined using a pancake coil probe. Tubes showing flaw indications

with a bobbin voltage greater than 3.6 volts will be plugged or

repaired.

Date of issuance: February 3, 1995.

Effective date: February 3, 1995.

Amendment No: 184.

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

Technical Specifications.

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

42337). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 3, 1995.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

No. 1, Pope County, Arkansas

Date of amendment request: November 8, 1994.

Brief description of amendment: The amendment revised the technical

specification section that describes the frequency for performing the

containment integrated leak rate tests.

Date of issuance: February 6, 1995.

Effective date: February 6, 1995.

Amendment No.: 175.

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

Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995, (60 FR

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

in a Safety Evaluation dated February 6, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: July 25, 1994.

Brief description of amendment: This amendment will upgrade

Technical Specification 3/4.7.1.6 for the Main Feedwater Line Isolation

Valves to be consistent with NUREG-1432, Standard Technical

Specifications for Combustion Engineering Plants.

Date of Issuance: February 9, 1995.

Effective Date: February 9, 1995.

Amendment No.: 71.

Facility Operating License No. NPF-16: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45024) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 9, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of application for amendments: July 25, 1994.

Brief description of amendments: These amendments implement GL 93-

05 Items 5.8, 6.1, 7.1 and 7.5.

Date of Issuance: February 9, 1995.

Effective Date: February 9, 1995.

Amendment Nos.: 133 and 72.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45023) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 9, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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: November 7, 1994.

Brief description of amendments: The amendments permit both

containment personnel airlock doors to be open while moving fuel during

refueling operations.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 1--Amendment No. 69; Unit 2--Amendment No. 58.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 7, 1994 (59 FR

63123). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 9, 1995.

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, Texas 77488.

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: November 8, 1994.

Brief description of amendments: The amendments permit the

substitution of an extended range neutron flux monitor for one of the

source range neutron flux monitors during refueling operations.

Date of issuance: February 13, 1995.

Effective date: February 13, 1995.

Amendment Nos.: Unit 1--Amendment No. 70; Unit 2--Amendment No. 59.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 7, 1994 (59 FR

63124). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 13, 1995.

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, Texas 77488.

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: June 6, 1994, as supplemented by letters

dated November 17, 1994, and December 5, 1994.

Brief description of amendments: The amendments modify Technical

[[Page 11146]] Specification 3/4.8.1.1, ``A.C. Sources'' by revising

the action statements and surveillance requirements for testing of the

standby diesel generators (SDGs). The amendments eliminate excessive

and unnecessary testing of the SDGs consistent with the guidance

provided in NUREG-1366, ``Improvements to Technical Specifications

Surveillance Requirements,'' NUREG-1431, ``Standard Technical

Specifications for Westinghouse Plants,'' Generic Letter 84-15,

``Proposed Staff Actions to Improve and Maintain Diesel Generator

Reliability,'' and Generic Letter 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operation.'' The changes include: (1) eliminating

the requirement to demonstrate the operability of an operable SDG

whenever an offsite AC power source is determined to be inoperable, or

whenever a support system or an independently testable component of

another SDG is inoperable, (2) eliminating the requirement to load the

diesel in 10 minutes during testing, (3) replacing the minimum required

loading for testing with a load band, (4) relocating some surveillance

requirements to the Diesel Fuel Oil Testing Program, and (5)

eliminating unnecessary loss-of-offsite power tests.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995, to be implemented within 60 days

of issuance.

Amendment Nos.: Unit 1--Amendment No. 68; Unit 2--Amendment No. 57.

Facility Operating License Nos. NPF-76 and NPF-80. Amendment

revised the Technical Specifications.

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

37073). The November 17, 1994, and December 5, 1994, submittals

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 2, 1995.

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, Texas 77488.

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

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

Date of application for amendments: August 3, 1994.

Brief description of amendments: The amendments relocate the

Radiological Effluent Technical Specifications to other controlled

documents consistent with NRC Generic Letter 89-01.

Date of issuance: February 10, 1995.

Effective date: February 10, 1995.

Amendment Nos.: 189 and 175.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55873).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 10, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: October 24, 1994, as

supplemented by letter dated December 16, 1994.

Brief description of amendment: This amendment modifies Technical

Specifications Table 4.1

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