Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from August 18, 1995, through August 30, 1995.

The last biweekly notice was published on Wednesday, August 30, 1995

(60 FR 45172).

[[Page 47614]]

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 October 13, 1995, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

[[Page 47615]]

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

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

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

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

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: July 17, 1995

Description of amendment request: The requested change to Technical

Specification (TS) section 3.8 would specify that the spent fuel

building refueling filter fan and at least one containment purge fan

shall be shown to operate within plus or minus 10 percent of the design

flow.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:The proposed change to TS is to

revise Section 3.8.2.c. This TS section currently states ``All filter

system fans shall be shown to operate within [plus or minus] 10% of the

design flow.'' The proposed requirements are as follows:

c.1 The Spent Fuel Building refueling filter fan shall be shown

to operate within [plus or minus] 10% of the design flow.

c.2 At least one Containment purge filter fan shall be shown to

operate within [plus or minus] 10% of the design flow and must be

operable during core alterations or movement of irradiated fuel

assemblies, or at least one automatic containment isolation valve in

each line penetrating the containment which provides a direct path

from the containment atmosphere to the outside atmosphere shall be

securely closed.

This proposed change does not involve a significant hazards

consideration for the following reasons.

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change clarifies the operating requirements

for the Containment purge and Spent Fuel Building refueling filter

systems. This proposed change to the TS specifically delineates the

fan filter systems required for refueling operations and does not

change the physical operation of the filter systems. The affected

systems are not involved in the initiation of any accident. The

system response to previously analyzed accidents, including system

flows and filter efficiencies will not be altered by the proposed

change. These changes are enhancements to clarify existing TS

requirements that will not increase the probability or consequences

of a previously analyzed accident.

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

or different kind of accident from any accident previously

evaluated. The proposed change merely clarifies the specific filter

systems that are necessary to mitigate a fuel handling accident

during core alterations or the movement of irradiated fuel

assemblies and is consistent with the accident analysis in Section

15.7.4 of the Updated Final Safety Analysis Report (UFSAR). This

proposed change does not involve the addition or modification of

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

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

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

different kind of accident from any accident previously evaluated.

3. The proposed change clarifies which filter systems that must

be capable of mitigating a design basis fuel handling accident

during core alterations or the movement of irradiated fuel

assemblies and is consistent with the accident analysis in Section

15.7.4 of the UFSAR. The proposed change will not result in an

increase in the Control Room or offsite radiation doses. The

performance of the filtration systems, including adsorption

efficiencies, will not change. Therefore, the proposed change does

not involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, SC 29550

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

& Light Company, Post Office Box 1551, Raleigh, NC 27602

NRC Project Director: David B. Matthews

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

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

Date of amendment request: June 30, 1995

Description of amendment request: The proposed amendments would

modify the emergency diesel generator testing requirements in the

Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of occurrence of any accident

previously evaluated.

The proposed changes to the Technical Specifications will change

the scope of EDG [Emergency Diesel Generator] testing that is

performed on a refueling cycle frequency. The proposed changes will

eliminate the requirement to perform sequenced loading of the EDG as

part of the hot restart test, and will allow the hot restart test to

be initiated from any EDG start signal. The revised requirements

will eliminate testing that is redundant, provides no additional

meaningful information, significantly constrains scheduling of

refueling outage maintenance and testing, and impacts the

availability of systems and components important to safety. The

proposed testing requirements satisfy the underlying purpose of the

EDG hot restart test. The testing in accordance with the proposed

requirements will verify the ability of each EDG to complete the

start up sequence from an equilibrium temperature immediately

following operation at full load for a period of time long enough to

stabilize operating temperature.

A two hour period for operation at full load has been chosen to

ensure that full load operating temperature has stabilized prior to

shutdown preceding the hot restart test. Momentary transients

outside the full load operating band of 3600 to 4000 kW will not

invalidate the two hour run since momentary transient will not

significantly affect operating temperature. Brief operation

subsequent to a momentary transient will normalize operating

temperature. Since the proposed changes impact only surveillance

requirements used to periodically verify the operability of a

required safety system, and since the proposed changes provide an

[[Page 47616]]

equivalent level of testing and eliminate redundant testing, the

proposed changes will not impact the operability or availability of

a required system.

Operation in accordance with the revised requirements will not

increase the likelihood that a transient initiating event will occur

since transients are initiated by equipment malfunction and/or

catastrophic system failure. The revised requirements affect testing

that is performed on a Refueling Cycle frequency. Testing in

accordance with the proposed requirements will not increase the

probability of failure of the EDGs since the testing will provide an

equivalent level of testing to verify the operability of the EDGs.

In addition, failure of an EDG to start or failure of an EDG while

operating is not assumed to be an initiating event of an accident

considered in the Updated Final Safety Analysis Report (UFSAR).

Based on the above, operation in accordance with the proposed

requirements will not significantly increase the probability of

occurrence of any accident previously evaluated.

The proposed requirements will meet the underlying purposed of

the existing testing requirements. The proposed testing will ensure

the ability of the EDG to start from a hot condition in the unlikely

event of an accident. The proposed changes will eliminate testing

requirements that are redundant and unnecessarily challenge the

reliability of the EDGs by requiring unnecessary wear and cycling of

the diesel engine and auxiliary systems. Since the proposed changes

will not adversely affect the operability or availability of the

EDGs, the ability of the EDGs to operate and power equipment

important to safety will not be impacted and the ability to mitigate

the consequences of accidents previously evaluated will not be

affected. Based on the preceding discussion, the consequences of

accidents previously evaluated will not significantly increase.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes to the Technical Specifications do not

involve the addition of any new or different types of safety related

equipment, nor do they involve the operation of equipment required

for safe operation of the facility in a manner different from those

addressed in the UFSAR. No safety related equipment or function will

be altered as a result of the proposed changes. Also, the procedures

that govern normal operation and recovery from an accident are not

affected by the proposed changes.

The proposed changes will eliminate testing requirements that

are redundant and provide no additional meaningful information.

Testing in accordance with the revised requirements will provide an

equivalent level of confidence in the reliability of the EDG systems

to complete the start up sequence from a hot condition. The proposed

testing requirements satisfy the purpose Regulatory Guide 1.108 in

that the testing requirements will ensure EDG operability and

reliability. In addition, the proposed changes are consistent with

the changes recommended by the NRC in Generic Letter 93-05. Since no

new failure modes or mechanisms are introduced by the proposed

changes, the possibility of a new or different kind of accident is

not created.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

Plant safety margins are established through LCOs, limiting

safety system settings, and safety limits specified in the Technical

Specifications. There will be no changes to either the physical

design of the plant or to any of these settings or limits as a

result of the proposed changes. The proposed changes will eliminate

testing requirements that are redundant and provide no additional

information. Testing in accordance with the revised requirements

will verify the ability of the EDGs to complete the start up

sequence from a hot condition as is intended by the recommended

testing in Regulatory Guide 1.108. In addition, the proposed changes

are consistent with the changes recommended by the NRC in Generic

Letter 93-05. Since the proposed changes will not impact the

availability or operability of the EDGs to perform their intended

function and since no LCOs, safety limits, or safety system settings

are affected by the proposed changes, there is no 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: Waukegan Public Library, 128

N. County Street, Waukegan, IL 60085

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

Austin, One First National Plaza, Chicago, IL 60603

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: July 26, 1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications to allow rod

misalignment of +/- 18 steps at or below 90% of rated thermal power. In

addition, a change is proposed to the Limiting Condition for Operation

range of rod travel from 228 to ``All Rods Out.'' The introduction of

``All Rods Out'' is consistent with Amendment 167/161 which approved

the removal of Technical Specification 3.1.3.6, ``Rod Insertion Limit''

from the Technical Specifications and placement into the Core Operating

Limits Report (COLR).

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 any accident previously evaluated.

The proposed limits on rod misalignment do not increase the

probability of an accident. The Technical Specifications' allowed

increase in peaking factor limits as power is reduced accommodates

an increase in rod misalignment of [plus or minus] 18 steps below

90% of RTP [rated thermal power]. The initial conditions remain

unchanged from that assumed in the Updated Final Safety Analysis

Report (UFSAR). Therefore, this proposed change poses no significant

increase in the probability or consequences of any 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.

No new accident scenarios, failure mechanisms or limiting single

failure are introduced as a result of implementing the proposed rod

misalignment criteria. The institution of the proposed rod

misalignment criteria will have no adverse effect, nor does it

challenge, the performance of any other safety related system.

Therefore, the proposed amendment does not in any way 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 the margin of

safety.

Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in the margin of

safety. The margin of safety, as defined in the BASES for the

Technical Specifications, is not significantly affected by the

changes to the rod misalignment limit. The Technical Specifications'

allowed increase in peaking factor limits as power is reduced

accommodates an increase in rod misalignment of [plus or minus] 18

steps below 90% of RTP. The initial conditions remain unchanged from

that assumed in the UFSAR. Since the peaking factor limits are not

modified, the proposed change does not constitute a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, FL 33199

[[Page 47617]]

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 26, 1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications to delete the

requirement to adjust the Nuclear Instrumentation System (NIS) downward

when operating at less than 70% of rated thermal power (RTP).

At reduced power levels (i.e., less than 70% of RTP), calorimetric

power measurement uncertainties are most influenced by the feedwater

flow measurements, which have the potential for large flow

uncertainties under low flow conditions. These calorimetric

uncertainties create the potential for a non-conservative gain

adjustment of the NIS when the NIS is adjusted downward to match

calorimetric power at reduced power levels, and may result in a non-

conservative NIS power level indication when operating at higher power

levels. Inappropriate gain adjustments could cause the NIS Power Range

High Neutron Flux trip to occur at power levels beyond that assumed in

the plant safety analyses. The proposed changes would correct this

situation.

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 change does not involve any physical changes to the

NIS. Implementation of the proposed change does not affect the

probability of failure of the NIS and does not alter the method in

which protection is afforded by the NIS for the reactor and primary

system. Therefore, the proposed change does not result in an

increase in the severity or consequences of any accident previously

evaluated.

The proposed change in Technical Specifications to remove the

requirement which could result in non-conservative gain adjustments

of the NIS at reduced power levels (below 70% of RTP), will have no

significant effect on the probability or consequences of licensing

basis events; and the probability or consequences of an accident

previously evaluated for Turkey Point has not been significantly

increased. Therefore, 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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed change does not result in a change in the method in

which the NIS provides plant protection. No change is being made

which alters the function of the NIS. Therefore, the proposed change

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

accident nor involve a reduction in a margin of safety as defined in

the Safety Analysis Report.

The change in Technical Specifications associated with the

removal of the requirement which could result in non-conservative

gain adjustments of the NIS at reduced power levels (below 70% of

RTP) will 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 the margin of

safety.

This change in Technical Specifications only affects the removal

of the requirement which has the potential for non-conservative gain

adjustments of the NIS at reduced power levels (below 70% of RTP);

these changes do not alter the manner in which protection is

afforded for the reactor and primary system. In addition, the

fundamental process for implementation of the calorimetric power/NIS

comparison remains the same.

The changes in Technical Specifications associated with the

removal of the requirement, which could lead to non-conservative

gain adjustments of the NIS at reduced power levels (below 70% of

RTP), 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, FL 33199

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 26, 1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) to incorporate

certain changes which are consistent with guidance provided by NUREG-

1366 and NRC Generic Letter (GL) 93-05, ``Line-Item Technical

Specification Improvements to Reduce Surveillance Requirements for

Testing During Power Operation.'' The following proposed changes are

requested:

(1) TS SR 4.1.3.1.2: Change the frequency interval for control rod

movement test from monthly to quarterly.

(2) TS SR 4.6.5.1: Change the hydrogen monitor calibration from

quarterly to each refueling interval, and the analog channel

operational test from monthly to quarterly.

(3) TS SR Table 4.3-3: Change the analog channel functional test

from monthly to quarterly for radiation monitors. Correct spelling of

'Radioactivity' in Item 1.a.

(4) TS SR 4.4.6.2.2: Increase the time allowed in COLD SHUTDOWN

before leak testing the Reactor Coolant System (RCS) isolation valves

is required, from 72 hours to 7 days.

(5) TS SR 4.10.1.2: Changes the requirement for a rod drop test

prior to reducing SHUTDOWN MARGIN from ``within 24 hours'' to ``within

7 days''.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed amendments conform to the guidance given in

Enclosure 1 of the NRC Generic Letter 93-05. The overall functional

capabilities of the rod control system, RCS pressure isolation

valves, the hydrogen monitoring system, and the radiation monitoring

systems will not be modified by the proposed change. These

amendments will not involve a significant increase in the

probability or consequences of an accident previously evaluated for

the following reasons:

(1) Increasing the interval of control rod movement testing will

reduce the possibility of testing-related reactor trips and dropped

rods, and result in fewer challenges to safety systems and plant

transients.

(2) Increasing the interval of hydrogen monitor calibration and

operational tests will result in a reduction in equipment

degradation and reduce a burdensome task on personnel resources.

[[Page 47618]]

(3) Increasing the interval of radiation monitor functional

tests will result in less equipment degradation as well as reducing

the potential for testing-related isolations of the control room,

fuel handling building, auxiliary buildings, and various process

lines.

(4) Increasing the time allowed in COLD SHUTDOWN prior to leak

testing RCS isolation valves will permit plant personnel to focus on

short notice outage recovery and minimize personnel radiation

exposure. Since the methods and the acceptance criteria used for the

leak test are not altered, increasing the time from 72 hours to 7

days will not significantly alter the associated risk.

(5) Increasing the time required to perform rod tests prior to

reducing the SHUTDOWN MARGIN will result in only one rod drop test

vice two following a refueling outage, which will in turn reduce

plant transients and personnel resource requirements.

(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 use of the proposed changes to the TS can not create the

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

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the surveillance interval changes and clarifications, since the

proposed changes do not involve the addition or modification of

equipment nor do they alter the design or operation of affected

plant systems.

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

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems are unchanged by the proposed amendments. The proposed

changes to the TS which establish new or clarify old surveillance

intervals consistent with the NRC Generic Letter 93-05 line-item

improvement guidance do not significantly reduce any of the margins

of safety even though the number of surveillances is decreased.

These requested amendments are justified by the following reasoning

from NUREG-1366:

(1) The surveillances could lead to plant transients which would

challenge safety systems unnecessarily as in the cases of control

rod movement tests and post-refueling rod drop tests.

(2) The surveillances result in the unnecessary wear to

equipment as in the cases of the hydrogen and radiation monitor

surveillances.

(3) The surveillance result in radiation exposure to plant

personnel which is not justified by the safety significance of the

surveillances as in the case of the time requirement for leak-

testing RCS isolation valves when in COLD SHUTDOWN.

(4) The surveillances place an unnecessary burden on plant

personnel because the time required is not justified by the safety

significance of the surveillance, i.e. hydrogen monitor and post-

refueling rod drop tests.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, FL 33199

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 26, 1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specification Administrative

Controls Section 6.9.1.7 to reflect the use of the Westinghouse NOTRUMP

model in the Small Break Loss-of-Coolant Accident (SBLOCA) analysis

used in determining the K(z) curve contained in the Core Operating

Limits Report (COLR). The following references would be added to

Section 6.9.1.7 (COLR) of the Administrative Controls section of Turkey

Point Units 3 and 4 TS: WCAP-10054-P-A, (proprietary) and

WCAP-10081-NP-A, (non-proprietary), ``Westinghouse Small Break ECCS

Evaluation Model Using the NOTRUMP Code'', October, 1985.'' WCAP-10054-

P-A Addendum 2, (proprietary), ``Addendum to the Westinghouse Small

Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection

into the Broken Loop and COSI Condensation Model'', August, 1994.''

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 modification to the current Section 6.9.1.7 of the

Administrative Controls section of the Turkey Point Technical

Specifications to include the references to WCAP-10054-P-A, ``Small

Break ECCS Evaluation Model Using the NOTRUMP Code'', and WCAP-

10054-P-A Addendum 2 for the COSI model, does not involve an

increase in the probability or consequences of an accident

previously evaluated. This modification to the Technical

Specification does not change the probability of occurrence

previously evaluated.

This change does not affect the integrity of the fission product

barriers utilized for mitigation of radiological dose consequences

as a result of an accident. The addition of the new methodology used

for Turkey Point uprating analysis does not change, degrade, or

prevent the response of safety related mitigation systems to

accident scenarios, as described in the Updated Final Safety

Analysis Report (UFSAR) Chapter 14. Therefore, the licensee

concluded that the probability or consequences of an accident

previously evaluated are not increased.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The modification to the current Section 6.9.1.7 of the

Administrative Controls section of the Turkey Point Technical

Specifications to include the references to WCAP-10054-P-A, ``Small

Break ECCS Evaluation Model Using the NOTRUMP Code'', and WCAP-

10054-P-A Addendum 2 for the COSI model, will not create the

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

previously evaluated. No new operating configuration is being

imposed by the addition of the references to the Technical

Specification. Therefore, no new failure modes or limiting single

failures have been identified. The licensee concludes that no new or

different kind of accidents from those previously evaluated have

been created as a result of this revision.

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

amendment would not involve a significant reduction in the margin of

safety.

The modification to the current Section 6.9.1.7 of the

Administrative Controls section of the Turkey Point Technical

Specifications to include the references for the Small Break ECCS

Evaluation Model Using the NOTRUMP Code will not involve a reduction

in the margin of safety. The SBLOCA analysis results show that the

limits of 10 CFR 50.46 are maintained as follows. The new calculated

value of worst-case PCT will be 1688 deg.F, which is less than the

limit of 2200 deg.F. There is significant margin in the current

SBLOCA analysis such that the total cladding oxidation limit of 17

percent will not be challenged. Further, the calculated total amount

of hydrogen generated has been determined to remain less than 1

percent. The SBLOCA hydraulic forces are not affected by the K(z)

curve and the licensee concludes that the core will remain amenable

to cooling. Additionally, post-LOCA long term core cooling and hot

leg switchover evaluations are not impacted by the K(z) curve.

Therefore, there is no significant reduction in the margin of

safety.

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

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

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

amendment request

[[Page 47619]]

involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, FL 33199

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 26, 1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications to achieve

consistency throughout these documents by (a) removing outdated

material, (b) incorporating administrative clarifications and

corrections, and (c) correcting typographical errors.

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 amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed amendments are purely administrative in nature.

These amendments will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because they do not affect assumptions contained in plant safety

analyses, the physical design and/or operation of the plant, nor do

they affect Technical Specifications that preserve safety analysis

assumptions. Therefore, the proposed changes do not affect the

probability or consequences of accidents previously analyzed.

(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 use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the administrative changes and clarifications, since the proposed

changes do not involve the addition or modification of equipment nor

do they alter the design or operation of affected plant systems,

structures, or components.

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

amendment would not involve a significant reduction in the margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. The modified specifications which correct administrative

errors and clarify existing Technical Specification requirements do

not significantly reduce any of the margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, FL 33199

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

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

NRC Project Director: David B. Matthews

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: August 17, 1995

Description of amendment request: The proposed amendment would

allow the containment to be opened after about 11 days following

shutdown during refueling and would redefine the operability

requirements for selected engineered safety feature systems such that

these systems are only required to be operable during the calculated

decay period. The proposed changes will not remove requirements for

systems to mitigate potential vessel draindown events, will not remove

requirements for systems required for decay heat removal, and will

continue to require high water level over the vessel during fuel

movement. Programs are in place to close the containment, if needed, to

address shutdown risk concerns.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed limits on recently irradiated fuel is used to

establish operational conditions where specific activities represent

situations where significant radioactive releases can be postulated.

These operational conditions are consistent with the design basis

analysis. Because the equipment affected by the revised operational

conditions is not considered an initiator to any previously analyzed

accident, inoperability of the equipment cannot increase the

probability of any previously evaluated accident.

The proposed applicability in conjunction with existing

administrative controls on light loads, bounds the conditions of the

current design basis fuel handling accident analysis. The analysis

also concludes the limiting offsite radiological consequences are

well within the acceptance criteria of NUREG 0800, Section 15.7.4

and GDC 19. The analysis is also conducted in a conservative manner

containing margins in the calculation of mechanical analysis, iodine

inventory and iodine decontamination factor. Each of these

conservatisms will further decrease the consequences. Therefore, the

proposed changes do not significantly increase the probability or

consequences of any previously evaluated accident.

The proposed limits are used to establish operational conditions

where specific activities represent situations where significant

radioactive releases can be postulated. In addition, the changes to

operation are consistent with previous limits -- only allowing

increased flexibility after the radiological consequences are

assured to remain within accepted limits. Therefore, these

operational conditions are consistent with the design basis

analysis. The proposed changes do not introduce any new modes of

plant operation and do not involve physical modifications to the

plant. Therefore, the proposed changes do not create the possibility

of a new or different kind of accident from any previous analyzed.

The revised limits are used to establish operational conditions

where specific activities represent situations where significant

radioactive release can be postulated. These operational conditions

are consistent with the design basis analysis and are established

such that the radiological consequences are at or below the current

RBS licensing limit. Safety margins and analytical conservatisms

have been evaluated and are well understood. Conservative methods of

analysis are maintained through the use of accepted methodology and

benchmarking the proposed methods to previous analysis. Margins are

retained to ensure that the analysis adequately bounds all

postulated event scenarios. The proposed change only eliminates some

excess conservatism from the analysis.

EOI has implemented NUMARC 91-06 guidelines for shutdown

operations at RBS. Shutdown Operations Protection Plan and Primary-

Secondary Containment Integrity procedures presently include

guidance for closure of the containment hatch and other significant

opening in containment, in addition to the requirements contained in

the license and design basis. This additional protection will

enhance the ability to limit offsite effects.

Acceptance limits for the fuel handling accident are provided in

10CFR100 with additional guidance provided in NUREG 0800, Section

15.7.4 Excess margin is the difference between the postulated doses

and the corresponding licensing limit. In the

[[Page 47620]]

initial review of River Bend Station for operation (NUREG-0989, Section

15.7.4), the NRC accepted the design and analysis based on meeting

the guideline dose limits of 10CFR100 and SRP 15.7.4. The proposed

applicability continues to ensure that the whole-body and thyroid

doses at the exclusion area and low population zone boundaries, as

well as control room doses, are below the corresponding licensing

limit. These margins are unchanged; 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: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

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

1400 L Street, N.W., Washington, DC 20005

NRC Project Director: William D. Beckner

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

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

Michigan

Date of amendment requests: June 20, 1995 (AEP:NRC:0692CX)

Description of amendment requests: The proposed amendments would

remove the requirements for fire protection systems from the licenses

and the Technical Specifications (T/S) in accordance with the

provisions and guidance of Generic Letters (GL) 86-10, ``Implementation

of Fire Protection Requirements,'' 88-12, ``Removal of Fire Protection

Requirements from Technical Specifications,'' and 93-07, Modification

of the Technical Specification Administrative Control Requirements for

Emergency and Security Plans.''

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:

We have evaluated the proposed T/S changes and have determined

that the changes should involve no significant hazards consideration

based on the criteria established in 10 CFR 50.92(c). Operation of

CNP [Cook Nuclear Plant] in accordance with the proposed amendment

will not satisfy any of the following criteria.

(a) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are administrative in nature, in that it

moves the T/Ss portion of the Fire Protection Program from the T/Ss

to the UFSAR [Updated Final Safety Analysis Report] and the

implementing procedures. This is accomplished by referencing in the

UFSAR and the documents which address the Fire Protection Program in

greater detail. Thus, the proposed changes will not revise the

requirements for fire protection equipment operability, testing, or

inspection, but only moves the T/Ss portion of the Fire Protection

Program to implementing procedures.

As fire protection requirements are only being relocated

following the guidance of GLs 86-10, 88-12, and 93-07, the proposed

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

accident from any previously analyzed.

The proposed changes do not involve any physical alteration of

plant configurations, changes to setpoints, or operating parameters.

[These] are administrative changes that retain the existing fire

protection requirements and relocate these requirements from the T/S

to the UFSAR; therefore, these changes do not create the possibility

of a new or different kind of accident.

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

The proposed changes follow guidance contained in GLs 86-10, 88-

12, and 93-07 for incorporating the Fire Protection Program into the

UFSAR. A license condition will be implemented that will require

that no changes can be made to the Fire Protection Program that will

adversely affect the ability to achieve or maintain safe shutdown in

the event of a fire without prior NRC approval. 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 requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: John N. Hannon

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

Millstone Nuclear Power Station, Unit No. 2, New London,

Connecticut

Date of amendment request: August 23, 1995

Description of amendment request: The proposed amendment would

revise Technical Specifications Section 3.8.1.1 and the Bases for

Section 3/4.8. The proposed amendment would extend the Allowed Outage

Time (AOT) for an Emergency Diesel Generator (EDG) from 72 hours to 7

days.

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:

... NNECO concludes that these changes do not involve a

significant hazards consideration since the proposed change

satisfies the criteria of 10 CFR 50.92(c). That is, the proposed

changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The EDGs supply backup power to the essential safety systems in

the event of a Loss of Normal (offsite) Power. EDGs are not accident

initiators. Therefore, this change does not involve an increase in

the probability of any accident previously evaluated.

Although the EDGs provide backup power to components that help

mitigate the consequences of accidents previously evaluated, the

extension in the AOT does not affect any of the assumptions used in

the deterministic evaluations of these accidents. Thus, this change

will not increase the consequences of any accident previously

analyzed.

The increase in the EDG AOT introduces the potential to increase

the risk to the public since a longer time window provides an

opportunity to perform additional preventive maintenance to the EDG

while the plant is on-line. However, the extended AOT, by itself,

does not necessarily increase risk. The increase in the risk depends

on the total time during which an EDG was out of service and the

other equipment that is concurrently out of service with the EDG.

The total risk increase due to the EDG being out of service will not

be significant since that risk increase is monitored and kept at

acceptable levels in accordance with the risk monitor program.

Based on the above, the proposal to extend the AOT for the EDGs

(Technical Specification 3.8.1) does not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

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

from any previously analyzed.

The proposed change to extend the AOT for the EDGs (Technical

Specification 3.8.1) does not alter the physical design,

configuration, or method of operation of the plant. Therefore, the

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

of accident from any previously analyzed.

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

The proposed change to extend the AOT for the EDGs (Technical

Specification 3.8.1) do not affect the Limiting Conditions for

Operations or their bases. As a result, the deterministic analyses

performed to establish the margin of safety are unaffected. Thus,

the change does not involve a significant reduction in the margin of

safety.

[[Page 47621]]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

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

Millstone Nuclear Power Station, Unit No. 2, New London,

Connecticut

Date of amendment request: August 23, 1995

Description of amendment request: The proposed amendment would

extend the Allowed Outage Time (AOT) for an inoperable Safety Injection

Tank (SIT) from 1 hour to 24 hours, unless the SIT is inoperable due to

either boron concentration not within its limits or an inoperable level

or pressure instrument. For these two special cases, the proposed

change extends the AOT for an inoperable SIT to 72 hours. In addition,

the proposed amendment clarifies the completion times and conditions

for action statements and the criteria for surveillance requirements.

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:

... NNECO concludes that these changes do not involve a

significant hazards consideration since the proposed change

satisfies the criteria in 10 CFR 50.92(c). That is, the proposed

changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

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

Emergency Core Cooling System that mitigate the consequences of a

Loss of Coolant Accident (LOCA). As such, the SITs are not accident

initiators. Therefore, this change does not involve an increase in

the probability of any accident previously evaluated.

The increase in the AOT has the potential to increase the risk

if it becomes necessary to stay on-line longer than one (1) hour

with an inoperable SIT. However, the estimated risk impact is

negligible.

The SITs inject borated water into the reactor vessel (via the

cold legs) during the blowdown phase of a large break LOCA. The

introduction of the inventory of borated water from all four (4)

SITs is needed to ensure adequate reflooding of the core (i.e.,

minimize core damage) until the Engineered Safety Feature (ESF)

pumps can provide adequate core cooling. The SITs also provide

makeup water for the Reactor Coolant System (RCS) for smaller break

LOCAs. The extension of the AOT does not affect any of the

assumptions used in the deterministic evaluations of these

accidents. Thus, this change will not increase the consequences of

any accident previously evaluated.

The increased AOT extension to 72 hours, based solely on

instrumentation (level and pressure) malfunction, also does not

involve a significant increase in the consequences of an accident

previously evaluated as endorsed by the NRC in NUREG-1366,

``Improvements to Technical Specifications Surveillance

Requirements.''

The modification to the completion times and the modification of

the Surveillance Requirements for volumetric changes in the SIT as a

result of addition from the Refueling Water Storage Tank (RWST) also

do not involve a significant increase in the consequences of any

accident previously evaluated by the NRC in NUREG-1432, ``Standard

Technical Specifications for Combustion Engineering Plants.''

Based on the above, the proposed changes to extend the AOT for

an inoperable SIT, clarify action statements, and modify the

criteria for surveillance requirements, do not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

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

from any previously analyzed.

The proposed changes to extend the AOT for an inoperable SIT,

clarify action statements, and modify the criteria for surveillance

requirements, do not alter the physical design, configuration, or

method of operation of the plant. Therefore, the proposal does not

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

any previously analyzed.

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

The proposed changes to extend the AOT for an inoperable SIT,

clarify action statements, and modify the criteria for surveillance

requirements, do not affect the Limiting Conditions for Operations

(LCOs) of the SITs or the bases of the LCOs. As a result, the

deterministic analyses performed to establish the margin of safety

are unaffected. Thus, the change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

Northern States Power Company, Docket Nos. 50-282 and 50-306,

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

County, Minnesota

Date of amendment requests: May 4, 1995.

Description of amendment requests: The proposed amendments would

revise the pressurizer and main steam safety valve lift setting

tolerances from plus or minus 1% to plus or minus 3%, revise the Safety

Limit curves and revise the Technical Specification Section 2 to

conform to Standard Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated

The proposed changes increase the ``as-found'' setpoint

tolerances for the Pressurizer Safety Valves and Main Steam Safety

Valves from [plus or minus] 1% to [plus or minus] 3%. The proposed

changes do not involve any hardware modifications to plant

structures, systems, or components. Analyses have determined that

the proposed changes do not significantly affect the structural

integrity of either the Reactor Coolant System or the Main Steam

system.

The proposed setpoint tolerance of [plus or minus] 3% was

included in the assumptions for the performance of the reload safety

evaluations for the current fuel cycles, PI1-17 and PI2-16, and

subsequent Prairie Island fuel cycle analyses. These analyses

concluded that the minimum acceptable DNBR [departure from nucleate

boiling ratio] is maintained, over-pressure protection is

maintained, LOCA [loss-of-coolant accident] acceptance criteria are

met and offsite dose limits are not exceeded. These changes are

consistent with the guidance provided by Section III and XI of the

ASME [American Society of Mechanical Engineers] Code and Standard

Technical Specifications.

The proposed change to Technical Specification Figure TS.2.1-1

does not affect any existing accident analyses. This revision

ensures that the design bases and safety limits are accurately and

appropriately reflected in the Technical Specifications and will

ensure that plant operations are properly evaluated for DNBR

encroachment.

Therefore, the probability or consequences of an accident

previously evaluated are not affected by any of the proposed

amendments.

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

new of different

[[Page 47622]]

kind of accident from any accident previously analyzed The lift

setpoint the Pressurizer Safety Valves and Main Steam Safety Valves

will be restored to [plus or minus] 1% following testing, thus the

``as-left'' setpoint tolerance for the Pressurizer Safety Valves and

Main Steam Safety Valves are unchanged. Evaluations of plant normal

operation, transient and accident conditions have been performed

assuming these safety valve lift settings are [plus or minus] 3% of

the nominal setpoint and demonstrated that new or different kinds of

accidents are not created by the proposed changes.

The proposed changes to Technical Specification Figure TS.2.1-1

do not affect the design, function or operation of any Prairie

Island structures, systems or components. The curves show the loci

of points of reactor core differential temperature (an indication of

thermal power), Reactor Coolant System pressure, and average

temperature for which the minimum DNBR is not less than the safety

analysis limit, that fuel centerline temperature remains below

melting, that the average enthalpy in the hot leg is less than or

equal to enthalpy of saturated liquid, or that the exit quality is

within the limits defined by the applicable DNBR correlation. There

are no new failure modes introduced by the proposed changes to the

Figure. The changes conservatively adjust Figure TS.2.1-1 to current

plant conditions and ensure that the design is accurately reflected

and that the plant is operated in accordance with its design bases.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be created

[by] these amendments.

3. The proposed amendment will not involve a significant

reduction in the margin of safety

The proposed changes to the safety valve lift setting tolerances

are consistent with the guidance provided by Section III and XI of

the ASME Code and Standard Technical Specifications. Analyses have

demonstrated these valves will continue to perform their function of

protecting their respective system from over-pressurization under

all postulated transients and accidents. The changed setting

tolerances do not cause a reduction in any other safety margin such

as DNBR. SAFETY LIMIT curves are provided to define minimum

allowable safety margin for plant steady state operation, normal

operational transients and anticipated operational occurrences. The

SAFETY LIMITs represent a design requirement for establishment of

many of the RPS [reactor protection system] trip setpoints which

prevent reactor conditions from approaching the SAFETY LIMITs. The

proposed revision of the SAFETY LIMIT curves provide the minimum

safety margins with somewhat more conservatism than previously

included. No RPS trips setpoints are changed.

Therefore, a significant reduction in the margin of safety would

not be involved with these amendments.

Based on the evaluation described above, and pursuant to 10 CFR

Part 50, Section 50.91, Northern States Power Company has determined

that operation [of] the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by Nuclear

Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis, MN

55401

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: John N. Hannon

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

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

Pennsylvania

Date of amendment request: July 28, 1995

Description of amendment request: The proposed amendment would

revise the 250 volt DC [direct current] profiles in Technical

Specifications Surveillance 4.8.2.1 (d) (2c) to reflect the new load

profile calculations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

[Final Safety Analysis Report] FSAR Section 8.3 states that the

station batteries have sufficient capacity without the charger to

independently supply the required loads for four hours. The

Technical Specifications require that the batteries be surveilled to

dummy loads which are greater than the design loads. The load

profiles for the 250 VDC batteries were recalculated using discrete

increments of time when the loads would be in use for each of five

design basis events. The Technical Specification load profiles are a

composite of the worst case loads for the events plus margin. The

required ampere-hours for each battery using the new load profiles

is less than the ampere-hours required using the existing load

profiles. Therefore, since the load profiles envelop the actual

loads on the batteries, the change to the 250 VDC battery load

profiles does not involve a significant increase in the probability

or consequence of an accident previously evaluated.

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

from any accident previously evaluated.

As stated above, the 250 VDC batteries have sufficient capacity

to power the actual battery loads thus enabling them to perform

their intended function. Any postulated accident resulting from this

change is bounded by previous analysis. Therefore, the change to the

250 VDC battery load profiles does 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 Class 1E 250 VDC batteries are required to have sufficient

capacity and capability to ensure sufficient power is available to

supply the safety related equipment for (1) the safe shutdown of the

facilities and (2) the mitigation and control of accident conditions

within the facilities. The proposed load profiles envelope the worst

case loads plus margin.

The ampere-hours removed from the Class 1E 250 VDC batteries are

less for the proposed load profiles than the existing load profiles.

The ampere-hours available in the batteries after the batteries have

supplies[d] the emergency loads for 4 hours are: [See table in

subject application].

* * * * * * *

Engineering calculation shows that the Class 1E 250 VDC

batteries maintain at least 210 VDC at the Class 1E 250 VDC MCCs

while supplying the proposed loads, corrected for temperature and

aging. Since the Class 1E 250 VDC circuits are designed to operate

properly with a minimum of 210 VDC at the Class 1E MCCs, all the

Class 1E emergency equipment supplied from the Class 1E batteries

have sufficient voltage to operate for 4 hours after the loss of ac

power.

The Class 1E 250 VDC batteries and Class 1E 250 VDC battery

chargers have been sized using the proposed load profiles. The

Engineering calculation shows that the 120 cell, 12 positive plates

per cell battery banks are sufficient to supply the proposed load

profiles, corrected for temperature and aging. The same calculation

also shows that the Class 1E 250 VDC battery chargers have

sufficient capacity to re-charge the batteries from the proposed

emergency discharged conditions to the fully charged condition in 12

hours while continuing to supply the plant normal continuous loads.

Base upon the above discussion, the proposed changes to the

Technical Specification load profiles do not reduce the margin of

safety as defined 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: Osterhout Free Library,

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

Pennsylvania 18701

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

and

[[Page 47623]]

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

NRC Project Director: John F. Stolz

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

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania

Date of amendment request: August 11, 1995

Description of amendment request: The proposed amendment would

revise Susquehanna Unit 2 Technical Specification Table 3.3.7.5-1 as

follows:a.

Revise Item 13, Required Number of Channels from 1 to 2;b.

Revise Item 13, Minimum Channel Operable from 0 to 1;c.

Delete Footnote .

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:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Reestablishing the channel operability values in Item

13 of Technical Specification Table 3.3.7.5-1, and deleting

footnote , has no impact on the

probability or consequences of an accident previously evaluated. The

proposed change in the channel operability values is a return to the

values which were reviewed as part of the licensing basis.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Reestablishing the channel operability values in Item

13 of Technical Specification Table 3.3.7.5-1, and deleting

footnote , does not create the

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

previously evaluated. The change in the channel operability values

increases the required number of channels available for accident

monitoring. There is no correlation between increasing the number of

neutron flux accident monitoring channels available and the creation

of accident scenarios.

III. This change does not involve a significant reduction in a

margin of safety.

Reestablishing the channel operability values in Item

13 of Technical Specification Table 3.3.7.5-1, and deleting

footnote , does not involve a reduction

in a margin of safety. The proposed change increases the number of

required channels from current levels, and restores the values to

those which have historically been required. At the present time,

the number of required channels is being administratively controlled

at the proposed levels to ensure conservatism in operability.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701

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

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

NRC Project Director: John F. Stolz

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: May 12, 1995

Description of amendment request: The proposed change would extend

the surveillance test intervals for the emergency service water (ESW)

system to support 24 month operating cycles.

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 changes increase the interval between ESW system

surveillance tests. These changes are consistent with the guidance

provided in Generic Letter 91-04. These changes do not involve any

physical changes to the plant, nor do they alter the typical way the

ESW system functions. On-line testing will continue to assure

equipment availability. The type of testing and the corrective

actions required if the subject ESW surveillances fail remain the

same. As such, the proposed changes create no new impacts on

accidents previously evaluated.

Therefore, the proposed changes 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.

The proposed changes increase the interval between ESW system

surveillance tests. These changes are consistent with the guidance

provided in Generic Letter 91-04. The proposed changes do not change

the ability of the ESW system to provide heat removal for the ECCS

[emergency core cooling system] components and other equipment

essential to reactor shutdown. Past equipment performance and on-

line testing indicate the longer test intervals will not degrade ESW

equipment. No changes are proposed to the type of testing performed,

only to the length of the surveillance interval. The proposed

changes do not modify the design or operation of plant equipment,

therefore, no new or different failure modes are introduced.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The proposed changes increase the interval between ESW system

surveillance tests. These changes are consistent with the guidance

provided in Generic Letter 91-04. The proposed changes do not alter

the configuration of the ESW system nor change the manner in which

the ESW equipment functions. Past equipment performance and on-line

testing indicate the longer test intervals will not degrade ESW

equipment. Operation of the plant remains unchanged by the proposed

changes.

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, NY

13126

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

York, NY 10019

NRC Project Director: Ledyard B. Marsh

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 15, 1995

Description of amendment request: The proposed change would extend

the surveillance test intervals for the control rod system to support

24 month operating cycles.

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:

11. involve a significant increase in the probability or consequences

of an accident previously evaluated.

[[Page 47624]]

The proposed changes increase the interval between control rod

system surveillance tests. These changes are consistent with the

guidance provided in Generic Letter 91-04. These changes do not

involve any physical changes to the plant, nor do they alter the way

the control rod system functions. The type of testing and the

corrective actions required if the subject control rod surveillances

fail remain the same. As such, the proposed changes create no new

impacts on accidents previously evaluated.

The reactivity margin - core loading test can be safely extended

to accommodate the 24 month operating cycle. The calculation of

reactivity margin takes into account the longer operating cycle.

The control rod scram time test can be safely extended to

accommodate a 24 month operating cycle. Operating experience has

indicated that control rod scram times do not significantly change

over an operating cycle. Additional on-line testing provides

adequate assurance of equipment operability.

The SDIV [Scram Discharge Instrument Volume] vent and drain

valve operability test can be safely extended to accommodate a 24

month operating cycle. Evaluation of past surveillance performance

and additional on-line testing assure valve operability. The

operability of the mode switch and the reset switch is demonstrated

during shutdowns.

Therefore, the proposed changes do not involve a significant

increase in the probability and do not change the consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes increase the interval between control rod

system surveillance tests. These changes are consistent with the

guidance provided in Generic Letter 91-04. The proposed changes do

not change the ability of the control rod system to provide rapid

reactivity control in order that no fuel damage results from any

abnormal operating transient. Past equipment performance and on-line

testing indicate the longer test intervals will not degrade control

equipment. No changes are proposed to the type of testing performed,

only to the surveillance interval length. The proposed changes do

not modify the design or operation of plant equipment, therefore, no

new or different failure modes are introduced.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The proposed changes increase the interval between control rod

system surveillance tests. These changes are consistent with the

guidance provided in Generic Letter 91-04. The proposed changes do

not alter the configuration of the control rod system nor change the

manner in which the control rod system functions. Past equipment

performance and on-line testing indicate the longer test intervals

will not degrade control rod equipment. Operation of the plant

remains unchanged by the proposed changes.

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, NY

13126

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

York, NY 10019

NRC Project Director: Ledyard B. Marsh

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: July 21, 1995

Description of amendment request: The proposed changes would

replace the title-specific list of members on the Plant Operating

Review Committee (PORC) with a more general statement of membership

requirements, similar to that used to define Safety Review Committee

membership; expand the scope of disciplines represented on the PORC to

include Nuclear Licensing and Quality Assurance; change several

management position titles; and, make several editorial corrections to

the Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Replacing the title specific list of PORC members with a

statement of membership requirements for the committee does not

reduce the effectiveness of the committee to advise the Resident

Manager (Site Executive Officer) on matters regarding nuclear

safety.

The proposed title changes for the Chief Nuclear Officer, Site

Executive Officer, Shift Manager, and Control Room Supervisor are

changes in title only and do not affect the responsibilities,

authority, qualification requirements, or reporting relationships of

these positions.

The change proposed for Specification 6.12 is administrative in

nature, reflecting a change previously approved elsewhere in

Technical Specifications.

The Radiological and Environmental Services Manager title change

proposed for Specification 6.11(A)2 is administrative in nature,

reflecting a change previously approved elsewhere in Technical

Specifications.

The remainder of proposed changes correct grammar or improve

consistency in Technical Specification formatting and do not affect

the meaning or intent of the specifications involved.

Operation of the James A. FitzPatrick Nuclear Power Plant in

accordance with the proposed amendment would not involve a

significant hazards consideration as defined in 10 CFR 50.92. The

changes are administrative in nature and would not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated,

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

from those previously evaluated, or

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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, NY 10019

NRC Project Director: Ledyard B. Marsh

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 27, 1995

Description of amendment request: The proposed change to the

Technical Specifications (TS) would incorporate updated pressure vs.

temperature operating limit curves contained in TS Figure 3.4.6.1-1 and

revise TS Surveillance Requirement 4.4.6.1.3 based on implementation of

Regulatory Guide 1.99, Rev. 2 in accordance with Generic Letter 88-11.

The changes are a result of data obtained from the first set of

specimen capsules removed during Refueling Outage 5.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident [...] previously evaluated.

The proposed changes assure that the existing safety limits are

not exceeded due to changing Reactor Vessel conditions. These

changes reflect the latest material testing

[[Page 47625]]

results in accordance with 10CFR50, Appendix G. The proposed changes to

the pressure and temperature limits do not increase the probability

of nonductile failures. The proposed changes to the surveillance

requirement and the associated changes to the Bases to include a

commitment to the methodology of Regulatory Guide 1.99, Rev. 2

ensures that the most limiting Reactor Vessel material is used in

the determination of the pressure-temperature operating limits.

Therefore, it may be concluded that the proposed changes do not

involve a significant increase in the probability or consequences of

an accident or malfunction of equipment important to safety

previously evaluated.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

No physical plant modifications or new operating configurations

result from these changes. These changes do not adversely affect the

design or operation of any system or component important to safety,

rather they establish limits to assure that operations remain within

acceptable safety boundaries.

Therefore, the possibility of a new or different kind of

accident from any previously evaluated will not be created.

3. Will not involve a significant reduction in a margin of

safety. Analysis of the capsule specimens has concluded that the

Reactor Vessel has sufficient fracture toughness for continued safe

operation, provided that operation remains within acceptable

pressure-temperature limits. The revised pressure-temperature curves

define these acceptable pressure-temperature limits during plant

operation. The proposed changes maintain the existing margins of

safety by modifying the operating limits based on the most limiting

of the actual reference temperature shifts. This new limit

considered analytical results of the capsule specimens, or a

predicted shift considering the most limiting pressure vessel

material. Changes to the Surveillance Requirement criteria and the

associated Bases to include a commitment to the methodology

contained in Regulatory Guide 1.99, Rev. 2 will ensure that the most

limiting plate or beltline weld material will be utilized in the

determination of the pressure-temperature limits.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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

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

NRC Project Director: John F. Stolz

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

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

San Diego County, California

Date of amendment requests: August 1, 1995

Description of amendment requests: The amendment request proposes

to change Technical Specification (TS) 3/4.3.2, Table 3.3-3,

``Engineered Safety Features Actuation System Instrumentation.'' TS 3/

4.3.2 includes the requirements for the minimum number of toxic gas

isolation system (TGIS) trains operable. The TS change request is to

extend the allowed TGIS outage times during replacement of TGIS

instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Toxic Gas Isolation System (TGIS) is designed to monitor and

mitigate the effects of toxic gas releases on control room

habitability. TGIS unavailability is not a precursor to any accident

previously evaluated in Chapter 15 of the San Onofre Updated Final

Safety Analysis Report (UFSAR). A risk assessment of the TGIS

instrumentation replacement activity was performed and found that

the likelihood of a loss of control room habitability beyond that

permitted by the Technical Specifications (TS) will not exceed 1E-6

over the duration of this TS change. In addition, a loss of control

room habitability does not necessarily lead to an accident or core

damage event. However, if a loss of control room habitability was

conservatively assumed to lead to a core damage event, this increase

in risk would still not constitute a significant increase in the

consequences or probability of any accident previously evaluated

since the increase is less than 3% of the average annual core damage

risk from internal events as reported in the San Onofre Individual

Plant Examination. Therefore, operation of the facility in

accordance with this proposed change does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This change extends the allowed outage times of the TGIS system.

The change does not affect the design or operation of any other

plant systems. An increase in TGIS unavailability is not a precursor

to any accident previously evaluated in Chapter 15 of the San Onofre

UFSAR. Therefore, operation of the facility in accordance with this

proposed change does not create the possibility of a new or

different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

During replacement of TGIS instrumentation a single channel of

TGIS will be maintained operable except during periods when

construction activity may result in spurious TGIS alarms. During

these periods the control room will normally be isolated except for

brief periods when the control room will be open to allow for air

exchange or to allow for CREACUS equipment repair. These periods,

when the control room is open without a TGIS channel available, will

not exceed 54 hours during the entire period when this change is in

effect. Operation with control room ventilation in the normal mode

with a single channel of TGIS operable for 44 days and no TGIS

channel available for up to 54 hours has been analyzed, and results

in an increase in the probability of a loss of control room

habitability which does not exceed 1E-6 over the duration of this TS

change. Therefore, this proposed change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Main Library, University of

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

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, CA 91770

NRC Project Director: William H. Bateman

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.

[[Page 47626]]

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

North Carolina

Date of application for amendment: October 24, 1994, as

supplemented July 21, 1995. The July 21, 1995, letter provides

clarification information and did not change the scope of the October

24, 1994, letter, or the initial no significant hazards consideration

determination.

Brief description of amendment: The proposed amendment would revise

the TS to allow the relocation of TS 3/4.3.7.12, Area Temperature

Monitoring; and the associated Bases in the TS to licensee-controlled

documents.

Date of issuance: August 28, 1995

Effective date: August 28, 1995

Amendment No.: 62

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

Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60379) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 28, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, NC 27605

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

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

Date of application for amendments: May 18, 1995, as supplemented

May 31, 1995

Brief description of amendments: The amendments revise the

frequency for conducting the Catawba Unit 2 Integrated Leak Rate Test

(ILRT) from a nominal frequency of once per 40 months to less than or

equal to 70 months. This also involves the granting of an exemption

from the requirements of 10 CFR Part 50, Appendix J, which is addressed

by separate correspondence.

Date of issuance: August 18, 1995

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

30 days from the date of issuance.

Amendment Nos.: 133 and 127

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32362) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 18, 1995.No significant

hazards consideration comments received: No

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

Black Street, Rock Hill, SC 29730

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley

Power Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: April 26, 1995

Brief description of amendment: This amendment adds a requirement

to Technical Specification (TS) 4.5.2.a to periodically verify that the

High Head Safety Injection (HHSI) pump minimum flow valve, 2CHS*MOV373,

is maintained open during plant operation in Modes 1, 2, and 3. Valve

2CHS*MOV373, must be maintained open to provide a minimum flowpath for

the HHSI pumps thereby minimizing the likelihood of HHSI pump damage

due to pump operation with insufficient flow. The amendment allows

flexibility for local verification of valve position or flow indication

if the control room indication is not available. Several editorial

changes to TS 3/4.5.2 are also being made to provide consistent format

with other TSs.

Date of issuance: August 25, 1995

Effective date: August 25, 1995

Amendment No.: 73

Facility Operating License No. NPF-73: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 6, 1995 (60 FR

29874). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 25, 1995. No significant

hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket No.

50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,

Georgia

Date of application for amendment: April 14, 1995, as supplemented

by letters dated June 22 and July 18, 1995

Brief description of amendment: The amendment eliminates response

time testing (RTT) requirements for selected sensors and specific loop

instrumentations for (1) the Reactor Protection System (RPS), (2) the

Isolation System, and (3) the Emergency Core Cooling System (ECCS). In

addition, the Note for Surveillance Requirement 3.3.6.1.7, which reads:

``Radiation detectors may be excluded,'' is being removed since RTT is

not required for any radiation detector that provides a primary

containment isolation signal as indicated in Table 3.3.6.1-1 of the TS.

Date of issuance: August 23, 1995

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

60 days from the date of issuance

Amendment No.: 137

Facility Operating License No. NPF-5: Amendment revised the

Technical Specifications.

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

35076) The June 22 and July 18, 1995, letters provided clarifying

information that did not change the scope of the April 14, 1995,

application and initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 23, 1995.No significant hazards

consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, GA 31513

[[Page 47627]]

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,

Burke County, Georgia

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

supplemented by letters dated June 14 and July 6, 1995.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) with editorial changes to the Action

Statements of TS 3.8.1.1 and 3.8.1.2 in order to reflect the

availability of a third offsite ac electrical source. Technical

Specification 4.8.1.1.1 is clarified to specify that the offsite ac

circuits connected to the onsite Class 1E distribution system are

required to be verified OPERABLE. A footnote is added to TS 3.8.3.1 to

allow the connection of the third offsite ac source to the onsite

busses.

Date of issuance: August 29, 1995

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

within 30 days from the date of issuance

Amendment Nos.: 90 and 68

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Technical Specifications.

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

6301) The June 14 and July 6, 1995, letters provided clarifying

information that did not change the scope of the January 3, 1995,

application and initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated August 29, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, GA 30830

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

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

Date of application for amendments: May 23, 1995

Brief description of amendments: The amendments revise the column

format for the Reactor Protection System and Engineered Safety Feature

Actuation System Setpoints

Date of issuance: August 24, 1995

Effective date: August 24, 1995

Amendment Nos.: 176 and 170Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32364) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 24, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, FL 33199

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: December 20, 1993, as

supplemented July 19, 1994, and February 28, 1995.

Brief description of amendments: The amendments revise the

surveillance requirements and load profiles for A, B, and N Train

batteries.

Date of issuance: August 22, 1995

Effective date: August 22, 1995

Amendment Nos.: 198 and 183

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

revised the Technical Specifications.

Date of initial notice in Federal Register: February 2, 1994 (59 FR

4939) and June 6, 1995 (60 FR 29879) The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

August 22, 1995.No significant hazards consideration comments received:

No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

Michigan

Date of application for amendments: May 25, 1995, and supplemented

June 30, 1995

Brief description of amendments: The amendments allow fuel

reconstitution when analyzed in accordance with NRC-approved

methodologies. The amendments are line item improvements based on NRC

Generic Letter 90-02, ``Alternative Requirements for Fuel Assemblies in

Design Features Section of Technical Specifications,'' supplement 1.

Date of issuance: August 22, 1995

Effective date: August 22, 1995

Amendment Nos.: 199 and 184

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

revised the Technical Specifications.

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

35081) The June 30, 1995, supplement provided a minor revision to the

proposed Technical Specification pages which was within the scope of

the original application and did not change the staff's initial

proposed no significant hazards considerations determination. The

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

Safety Evaluation dated August 22, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: February 14, 1995

Brief description of amendment: This amendment makes the following

administrative changes to the Maine Yankee (MY) Technical

Specifications (TS):

a. Removes responsibility for audits of the emergency and security

plans--including their implementing procedures--from the TS and assigns

that responsibility to the emergency and security plans,

b. Assigns review responsibility for significant, accidental,

unplanned, or uncontrolled radioactive releases to the Nuclear Safety

Audit and Review (NSAR) Committee,

c. Assigns additional reporting requirements to the NSAR Committee,

and

d. Provides the President of MY with the authority to initiate an

audit of any area of facility operation.

Date of issuance: August 22, 1995

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

within 30 days.

Amendment No.: 152

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

Technical Specifications.

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

16191) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 22, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Wiscasset Public Library, High

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

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

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

Date of application for amendment: May 18, 1995

Brief description of amendment: The amendment revises the minimum

temperature at which the reactor vessel head bolting studs are allowed

to be

[[Page 47628]]

placed under tension. In addition, the amendment revises the minimum

reactor vessel metal temperature during core critical operation,

revises the minimum reactor vessel metal temperature for pressure

tests, makes editorial changes, and revises the Bases for the

applicable section.

Date of issuance: August 23, 1995

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

immediately.

Amendment No.: 85

Facility Operating License No. DPR-21. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32369) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 23, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

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

Date of application for amendment: June 15, 1995

Brief description of amendment: The amendment changes the

definition for an alteration of the reactor core to one that is

consistent with the intent of the improved standard technical

specifications. The amendment also makes administrative changes to

several technical specification pages.

Date of issuance: August 28, 1995

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

30 days.

Amendment No.: 86

Facility Operating License No. DPR-21. Amendment revised the

Technical Specifications.

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

37097) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 28, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360

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

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

Connecticut

Date of application for amendment: April 28, 1995, as supplemented

August 2, 1995.

Brief description of amendment: The amendment changes Technical

Specification (TS) Sections 3.7.5, 4.7.5, and 3/4.7.5, to permit

Millstone Unit 3 to remain in operation with the average ultimate heat

sink water temperature greater than 75* F (but less than or equal to

77* F) for a period of 12 hours.

Date of issuance: August 28, 1995

Effective date: As of the date of issuance.

Amendment No.: 119

Facility Operating License No. NPF-49. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 6, 1995 (60 FR

29881). The information in the licensee's submittal of August 2, 1995,

did not require a change to the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated August 28, 1995.No

significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360

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

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

Obispo County, California

Date of application for amendments: June 29, 1995

Brief description of amendments: The amendments revise the combined

Technical Specifications (TS) for Diablo Canyon Nuclear Power Plant,

Unit Nos. 1 and 2 (DCPP) to add Mode 1 applicability to TS 3/4.4.2.2,

``Safety Valves - Operating,'' and changes the low-temperature

overpressure protection (LTOP) system enable temperature for Mode 4

applicability from 323 degrees F to 270 degrees F in TS 3/4.3.2.1,

``Safety Valves - Shutdown.''

Date of issuance: August 23, 1995

Effective date: August 23, 1995

Amendment Nos.: Unit 1 - Amendment No. 107; Unit 2 - Amendment No.

106

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

revised the Technical Specifications.

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

37098) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 23, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, CA 93407

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

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

Pennsylvania

Date of application for amendments: February 1, 1995, as

supplemented by letter dated June 20, 1995

Brief description of amendments: The requested changes would modify

the applicable operational conditions for the secondary containment

isolation radiation monitors located on the refueling floor and for the

monitor located in the railroad access shaft.

Date of issuance: August 24, 1995

Effective date: Both units, as of the date of issuance and is to be

implemented within 30 days

Amendment Nos.: 152 and 122

Facility Operating License Nos. NPF-14 and NPF-22. The amendments

revised the Technical Specifications.

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

16192). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 24, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

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

Pennsylvania

Date of application for amendments: August 31, 1994, as

supplemented by letters dated May 11, and July 3, 1995

Brief description of amendments: This amendment revises the

Technical Specifications to permit the relocation of the Turbine

Overspeed Protection System to the Updated Final Safety Analysis Report

and Controlled Plant Procedures.

Date of issuance: August 24, 1995

Effective date: August 24, 1995

Amendment Nos.: 100 and 64

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55884) The supplemental letters do not

[[Page 47629]]

change the initial no significant hazards consideration determination

nor the initial Federal Register notice. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

August 24, 1995.No significant hazards consideration comments received:

No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

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

Pennsylvania

Date of application for amendments: February 22, 1995

Brief description of amendments: The amendments revise the

Technical Specifications Surveillance Requirements to clarify the

Emergency Diesel Generator acceptable steady state voltage range.

Date of issuance: August 24, 1995

Effective date: As of the date of issuance and shall be implemented

within 30 days of issuance.

Amendment Nos.: 101 and 65

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20525) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 24, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: January 13, 1995

Brief description of amendment: The amendment revised the

Administrative Controls Section (6.0) of the Technical Specifications

for Hope Creek Generating Station to reflect organizational changes and

resultant management title changes.

Date of issuance: August 22, 1995

Effective date: August 22, 1995

Amendment No.: 77

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

Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32371) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 22, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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

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

County, New Jersey

Date of application for amendments: December 23, 1994

Brief description of amendments: The amendments to the Technical

Specifications revise the surveillance requirement to perform a visual

inspection of containment areas affected by containment entry when

containment integrity is established. They are consistent with Item 7.5

of Generic Letter 93-05, ``Line-Item Technical Specifications

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation.''

Date of issuance: August 24, 1995

Effective date: As of the date of issuance, to be implementd within

60 days.

Amendment Nos.: 174 and 155

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

revised the Technical Specifications.

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

6308) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 24, 1995.No significant

hazards consideration comments received: No

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

West Broadway, Salem, NJ 08079

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

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

San Diego County, California

Date of application for amendments: September 16, 1994

Brief description of amendments: These amendments revise Technical

Specification (TS) 3/4.2.1, ``Linear Heat Rate.'' The linear heat rate

(LHR) limit for steady state operation is revised from 13.9 kw/ft to

13.0 kw/ft. The Bases for TS 3/4.2.1, ``Linear Heat Rate,'' is also

being revised to reflect the new value.

Date of issuance: August 23, 1995

Effective date: August 23, 1995, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 2 - Amendment No. 124; Unit 3 - Amendment No.

113

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55892) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 23, 1995. No significant

hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

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

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

Date of application for amendments: May 3, 1995

Brief description of amendments: The amendments delay

implementation of Amendment Nos. 182 and 174 until implementation

problems are addressed. These changes revise the setpoints and time

delays for the auxiliary feedwater loss of power and the 6.9 kv

shutdown board loss of voltage and degraded voltage instrumentation.

Date of issuance: August 22, 1995

Effective date: August 22, 1995

Amendment Nos.: 207 and 197

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

revise the technical specifications.

Date of initial notice in Federal Register: May 23, 1995 (60 FR

27343) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 22, 1995.No significant

hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, TN 37402

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

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

Date of application for amendments: April 6, 1995 (TS 94-18)

Brief description of amendments: The amendments revise Surveillance

Requirement 4.0.5 by replacing the current Inservice Inspection program

and the Inservice Testing program requirements with the requirements

stated in the Standard Technical Specifications (NUREG-1431). The

amendments also delete Technical Specification 3/4.4.10, ``Structural

Integrity ASME Code Class 1, 2 and 3 Components,'' and its related

Bases information.

Date of issuance: August 22, 1995

Effective date: August 22, 1995

Amendment Nos.: 208 and 198

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

revise the technical specifications.

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20528)

[[Page 47630]]

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

Safety Evaluation dated August 22, 1995.No significant hazards

consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, TN 37402

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

Callaway County, Missouri

Date of application for amendment: April 17, 1995, as supplemented

on June 30, 1995

Brief description of amendment: The amendment revises Technical

Specifications Technical Specification 2.2.1, Table 2.2-1. The changes

address reducing repeated alarms and partial reactor trips by revising

the Overpower Delta-T setpoint function.

Date of issuance: August 21, 1995

Effective date: Immediately, to be implemented within 30 days.

Amendment No.: 102

Facility Operating License No. NPF-30. Amendment revises the

Technical Specifications.

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

24922). The June 30, 1995, letter provided supplemental information

that did not change the initial proposed no significant hazards

consideration determination.The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated August 21, 1995. No

significant hazards consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, MO 65251

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

Project No. 2, Benton County, Washington

Date of application for amendment: September 2, 1992

Brief description of amendment: The amendment revises the required

signal-to-noise ratio for the source range monitors, as recommended by

General Electric.

Date of issuance: August 23, 1995

Effective date: August 23, 1995

Amendment No.: 140

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

Technical Specifications.

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

37101) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 23, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, WA 99352

Dated at Rockville, Maryland, this 6th day of September 1995.

For the Nuclear Regulatory Commission

Jack W. Roe,

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

Reactor Regulation.

[Doc. 95-22616 Filed 9-12-95; 8:45 am]

BILLING CODE 7590-01-F