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

Federal RegisterNov 27, 1995

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from October 28, 1995, through November 9, 1995.

The last biweekly notice was published on Wednesday, November 8, 1995

(60 FR 56361).

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 December 27, 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

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limitations in the order granting leave to intervene, and have the

opportunity to participate fully in the conduct of the hearing,

including the opportunity to present evidence and cross-examine

witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: October 20, 1995.

Description of amendment request: The proposed one-time amendment

would revise the Calvert Cliffs Nuclear Power Plant, Unit No. 1, (CC-1)

Technical Specifications (TSs) by extending certain 18-month instrument

surveillance intervals by a maximum of 39 days to March 31, 1996. The

instruments involved are included in the reactor protective system,

engineered safety features actuation system, power-operated relief

valves, low-temperature overpressure protection system, remote shutdown

instruments, post-accident monitoring, radiation monitoring, and

containment sump level instruments.

The Commission issued Amendment No. 208 to Facility Operating

License No. DRP-53 and Amendment No. 186 to Facility Operating License

No. DRP-69 for the CC-1/2, respectively. The amendments permanently

extended the surveillance intervals for the instruments described above

from 18 months to 24 months after a specified number of the instruments

had been replaced. The amendments were effective immediately and to be

implemented on CC-2 within 30 days, but not implemented on CC-1 until

its restart after the spring 1996 refueling outage. All of the

instruments identified for replacement on CC-2 have been replaced, but

those identified for replacement on CC-1 have not been replaced, thus,

the reason for the later implementation date. The proposed one-time

amendment is needed prior to Amendment No. 208 being implemented

because of a change in the refueling schedule. The licensee has

provided technical justification to allow operation for an additional

short-time period of up to a maximum of 39 days.

CC-1 was initially scheduled to begin its refueling outage on

February 16, 1996, which would have been within the time frame

necessary to perform the required 18-month instrument surveillances

currently required for the instruments identified above. The licensee

has recently rescheduled the refueling outage for CC-1 to start March

15, 1996, several months after the initial amendment request and after

consultation with the Pennsylvania-New Jersey-Maryland power pool. The

revised schedule will allow the maximum use of the available fuel in

the CC-1 reactor core and will also allow the unit to operate for an

additional period of about 1 month during a period of potentially high

power demand. In addition, the delay will allow more time to plan and

prepare for the upcoming refueling outage. Performing the required

instrument surveillances at power would present an unwarranted

personnel safety risk and, in some cases, the surveillances cannot be

done during power operation because they would cause a unit trip. This

proposed one-time amendment will be superseded by Amendment No. 208

when it is implemented.

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. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed one-time change would extend 18-month instrument

surveillance intervals by a maximum of 39 days to March 31, 1996,

for specific Reactor Protective System (RPS), Engineered Safety

Features Actuation System (ESFAS), Power-Operated Relief Valve, Low

Temperature Overpressure Protection (LTOP), Remote Shutdown, Post-

Accident Monitoring (PAM), Radiation Monitoring, and Containment

Sump Level instruments.

The purpose of the RPS is to effect a rapid reactor shutdown if

any one or a combination of conditions deviates from a pre-selected

operating range. The system functions to protect the core and the

Reactor Coolant System (RCS) pressure boundary. The purpose of the

ESFAS is to actuate equipment which protects the public and plant

personnel from the accidental release of radioactive fission

products if an accident occurs, including a loss-of-coolant

accident, main steam line break, or loss of feedwater event. The

safety features function to localize, control, mitigate, and

terminate such incidents in order to minimize radiation exposure to

the general public. The PAM instruments provide the Control Room

operators with primary information necessary to take manual actions,

as necessary, in response to design basis events, and to verify

proper system response to plant conditions and operator actions. The

purpose of the Remote Shutdown System is to provide plant parameter

indications to operators on a Remote Shutdown Panel to be used while

placing and maintaining the plant in a safe shutdown condition in

the event the Control Room is uninhabitable. The indications are

used to verify proper system response to plant conditions and

operator actions. The LTOP System protects against RCS

overpressurization at low temperatures

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by a combination of administrative controls and hardware. Power-

Operated Relief Valves are set to lift before pressurizer safety

valves, and subsequently reseat to minimize the release of reactor

coolant from the RCS. The Containment Sump High Level Alarm System

provides an alarm in the Control Room to provide one of the

available indications of excessive RCS leakage during normal plant

operation. The Containment Area High Range Radiation Monitoring

System provides an indication of high radiation levels in

containment.

Failure of any of these systems is not an initiator for any

previously evaluated accident. Therefore, the proposed change would

not involve an increase in the probability of an accident previously

evaluated.

Surveillance and maintenance history has demonstrated good

capability for identifying adverse operation by individual

instruments. Baltimore Gas and Electric Company has the capability

to respond to an inoperable instrument by following the Technical

Specification Actions for an inoperable instrument or by performing

a channel calibration with the Unit at full power. However,

calibration of all the instruments at power is not desirable because

of personnel safety, personnel radiation protection goals, and plant

reliability concerns.

These factors provide assurance that the requested surveillance

extension will not adversely affect our ability to detect

degradation of the instruments. Also, either analysis is available

to show the instruments will operate properly during the requested

surveillance extension, or the surveillance program has shown that

problems will be identified and addressed appropriately. Therefore,

these channels will be able to perform the functions assumed in the

safety analysis, and there is no significant increase in the

consequences of an accident previously evaluated.

Therefore, the proposed Technical Specification changes do not

significantly increase the probability or consequences of an

accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

This requested increase in surveillance interval for RPS, ESFAS,

Power-Operated Relief Valve, LTOP, Remote Shutdown, PAM, Radiation

Monitoring, and Containment Sump Level instrument surveillances does

not involve a significant change in the design or operation of the

plant. No plant hardware is being modified as part of the proposed

change. The proposed change also does not involve any new or unusual

actions by plant operators. Therefore, this change would not create

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

accident previously evaluated.

3. Does operation of the facility in accordance with the

proposed amendment involve a significant reduction in a margin of

safety?

The RPS, ESFAS, Power-Operated Relief Valve, LTOP, Remote

Shutdown, PAM, Radiation Monitoring, and Containment Sump Level

instruments are designed to provide actuation signals and/or

indications to ensure appropriate action is taken in response to

design basis accidents. Channel checks, channel functional tests and

routine comparison of the redundant and independent parameter

indications provides a reliable indication of instrument operation.

Also, either analysis is available to show the instruments will

operate properly during the requested surveillance extension, or

instrument surveillance program has shown that problems will be

identified and addressed appropriately. During the requested

extension, these systems will be available to perform the functions

assumed in the Safety Analysis. Surveillance and maintenance history

have demonstrated good capability for identifying adverse operation

by individual instruments. Baltimore Gas and Electric Company has

the capability to respond to such adverse operation, including

performing channel calibrations at power. However, such work on all

the instruments is not desirable because of personnel safety,

personnel radiation protection goals, and plant reliability

concerns. Extending the surveillance interval provides additional

possibility for instrument components to malfunction by means such

as drift or instrument failure, which could allow plant parameters

to exceed design bases assumptions. We have determined that the

effect of the surveillance interval extension on safety is small,

and operation of the instruments in the extended interval would not

invalidate any assumption in the plant licensing basis.

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: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Ledyard B. Marsh.

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

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

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

Illinois

Date of amendment request: October 3, 1995.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TSs) for both stations to

implement 10 of the line item TS improvements recommended in Generic

Letter (GL) 93-05, ``Line-Item Technical Specifications Improvements to

Reduce Surveillance Requirements for Testing During Power Operation,''

dated September 27, 1993. The proposed changes also include editorial

changes on the affected TS pages.

The proposed changes from GL 93-05 are the following: (1) TS

4.1.3.1.2 (GL 93-05, Item 4.2), extending the interval for checking the

operability of each full-length rod not fully inserted in the core from

31 days to 92 days; (2) Table 4.3-3 (GL 93-05, Item 5.14), extending

the interval for the digital channel operational test for radiation

monitoring instrumentation in the table from monthly to quarterly; (3)

TS 4.4.3.2 (GL 93-05, Item 6.6), extending the interval between current

tests of the required groups of pressurizer heaters from 92 days to

each refueling outage; (4) TS 4.4.6.2.2.b (GL 93-05, Item 6.1),

extending the time the plant may be in cold shutdown before pressure

isolation valve testing is required, prior to entry into Operational

Mode 2, from 72 hours to 7 days; (5) TS 4.5.1.1.b (GL 93-05, Item 7.1),

revising the requirement to verify the boron concentration in an

accumulator within 6 hours of any volume increase to the accumulator

(greater than or equal to 70 gallons) so that the verification is not

required when the volume increase is from the refueling water storage

tank (RWST) and the RWST has not been diluted since verifying that the

boron concentration of the RWST is within the concentration limits for

the accumulators; (6) TS 4.6.2.1 (GL 93-05, Item 8.1), extending the

interval between tests to verify each containment spray nozzle is

unobstructed from 5 years to 10 years; (7) TS 4.6.4.1 (GL 93-05, Item

5.4), extending the interval for testing each hydrogen monitor for

combustible gas control from 31 days to 92 days for the analog channel

operational test, and from 92 days to each refueling outage for channel

calibration; (8) TS 4.6.4.2 (GL 93-05, Item 8.5), extending the

interval between tests to demonstrate operability of the hydrogen

recombiner system from 6 months to once each refueling outage; (9) TS

4.7.1.2.1.a (GL 93-05, Item 9.1), extending the interval between tests

of the auxiliary feedwater pumps from 31 days to 92 days on a staggered

test basis; and (10) TS 4.11.2.6 (GL 93-05, Item 13), extending the

interval for determining the quantity of radioactivity contained in

each gas decay tank, when radioactivity is being added to the tanks,

from 24 hours to 7 days, with the 24-hour frequency maintained during

the primary coolant degassing operation. The editorial changes are the

following: (1) TS 4.4.6.2.1.c, changes the word ``from'' to the word

``to,'' (2) TS 4.5.1.1.c, the

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change clarifies that the motor control center compartment is for each

accumulator isolation valve, (3) TS 4.5.1.2, deletes the footnote

because the operating cycle in the footnote is over for each unit, and

(4) TS 4.7.1.2.1.a.2 and 4.7.1.2.1.c, renumbers and rephrases (only TS

4.7.1.2.1.a.2) other surveillance requirements for the auxiliary

feedwater pumps because of the proposed change to TS 4.7.1.2.1.a to

implement GL 93-05, Item 9.1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The changes are consistent with GL 93-05 and NUREG-1366

[''Improvements to Technical Specifications Surveillance

Requirements,'' December 1992. In GL 93-05, the staff stated that it

concluded, in performing the study documented in NUREG-1366, that

safety can be improved, equipment degradation decreased, and an

unnecessary burden on licensee personnel eliminated by reducing the

frequency of certain testing required in the Technical

Specifications during power operation]. The changes eliminate

testing that is likely to cause transients or excessive wear of

equipment. An evaluation of these changes indicates that there will

be a benefit to plant safety. The evaluation, documented in NUREG-

1366, considered (1) unavailability of safety equipment due to

testing, (2) initiation of significant transients due to testing,

(3) actuation of engineered safety features that unnecessarily cycle

safety equipment, (4) importance to safety of that system or

component, (5) failure rate of that system or component, and (6)

effectiveness of the test in discovering the failure.

As a result of the decrease in the testing frequencies, the risk

of testing causing a transient and equipment degradation will be

decreased, and the reliability of the equipment will not be

significantly decreased.

The initial conditions and methodologies used in the accident

analyses remain unchanged. The proposed changes do not change or

alter the design assumptions for the systems or components used to

mitigate the consequences of an accident. Therefore, accident

analyses results are not impacted. Appropriate testing will continue

to assure that equipment and systems will be capable of performing

the intended function.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes either modify allowable intervals between

certain surveillance tests, delete surveillance requirements, or

alter an action statement with regard to the required testing. The

proposed changes do not affect the design or operation of any

system, structure, or component in the plant. The safety functions

of the related structures, systems, or components are not changed in

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

component reduced by the revised surveillance or testing

requirements.

Appropriate testing will continue to assure that the system is

capable of performing its intended function. The changes do not

affect the manner by which the facility is operated and do not

change any facility design feature, structure, system, or component.

No new or different type of equipment will be installed. Since there

is no change to the facility or operating procedures, and the safety

functions and reliability of structures, systems, or components are

not affected, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

All of the proposed technical specification changes are

compatible with plant operating experience and are consistent with

the guidance provided in GL 93-05 and NUREG-1366. The changes

eliminate unnecessary testing that increases the risk of transients

and equipment degradation. There is no impact on safety limits or

limiting safety system settings.

The remaining proposed changes are administrative in nature and

have no impact on the margin of safety of any technical

specification. They do not affect any plant safety parameters or

setpoints.

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: For Byron, the Byron Public

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

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

Wilmington, Illinois 60481.

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

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

NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. 50-373, LaSalle County

Station, Units 1, LaSalle County, Illinois

Date of amendment request: October 2, 1995

Description of amendment request: The proposed amendments would

revise Section 3.4.2 to change the safety/relief valve (SRV) safety

function lift setting tolerances from +1%, -3% to plus or minus 3% and

include as-left SRV safety function lift setting tolerances of plus or

minus 1%.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

the standards of 10 CFR 50.92(c). The NRC staff's review is presented

below.

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

the probability or consequences of an accident previously evaluated.

The probability of an accident previously evaluated will not

increase as a result of this change, because the only changes are the

tolerances for the SRV opening setpoints and the speed of the reactor

core isolation cooling system (RCIC) turbine and pump. Changing the

maximum allowable opening setpoint for the SRVs does not cause any

accident previously evaluated to occur, or degrade valve or system

performance in any way so as to cause an accident to occur with an

increased frequency. In addition, the increased speed of the RCIC

turbine and pump are within the design limits of the system. RCIC

operability and failure probabilities are not impacted by this change.

The consequences of an ASME Overpressurization Event are not

significantly increased and do not exceed the previously accepted

licensing criteria for this event. General Electric (GE) has calculated

the revised peak vessel pressure for LaSalle Station to be 1341 psig,

which is well below the 1375 psig criterion of the ASME Code for upset

conditions, referenced in Section 5.2.2, Overpressurization Protection,

of the Updated Final Safety Analysis Report (UFSAR), and NUREG-0519

(Safety Evaluation Report related to the operation of LaSalle County

Station, Units 1 and 2, March 1981), and Section 15.2-4, Closure of

Main Steam Isolation Valves (BWR) of NUREG-0800 (Standard Review Plan).

GE has also performed an analysis of the limiting Anticipated

Transient Without Scram (ATWS) event, which is the Main Steam Isolation

Valve (MSIV) Closure Event. This analysis calculated the peak vessel

pressure to be 1457 psig, which is sufficiently below the 1500 psig

criterion of the ASME Code for emergency conditions.

Per NUREG-0519, listed above, Section 5.4.1, and Technical

Specification 4.7.3.b, the RCIC pump is required to develop flow

greater than or

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equal to 600 gpm in the test flow path with a system head corresponding

to reactor vessel operating pressure when steam is supplied to the

turbine at 1000 +20, -80 psig. Increasing the turbine and pump speed

ensures these criteria will still be met and the consequences of an

accident will not increase.

Therefore, there is not a significant increase in the consequences

of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The only physical changes are to increase the allowable tolerances

for SRV opening setpoints and to increase the RCIC pump and turbine

speeds. These changes do not result in any changed component

interactions. The SRVs and RCIC will still provide the functions for

which they were designed. Since all of the other systems evaluated will

continue to function as intended, the proposed changes do not create

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

previously evaluated.

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

the margin of safety.

While the calculated peak vessel pressures for the ASME

Overpressurization Event and the MSIV closure ATWS Event are larger

than that previously calculated without the proposed setpoint tolerance

increases, the new peak pressures remain sufficiently below the

respective licensing acceptance limits associated with these events. In

addition, the actual L1C8 reload analysis of the ASME

Overpressurization Event will be verified to be within the licensing

acceptance limit for that event prior to Unit 1 Cycle 8 startup, as

required in the normal reload 10 CFR 50.59 process. These licensing

acceptance limits have been previously evaluated as providing a

sufficient margin of safety. For other accidents and transients, the

increased setpoint tolerances have a negligible effect on the results,

so the margin of safety is preserved.

The staff has reviewed the amendment request and the licensee's no

significant hazards consideration determination. Based on the review

and the above discussions, the staff proposes to determine that the

proposed changes do not involve a significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

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

NRC Project Director: Robert A. Capra.

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

County, Michigan

Date of amendment request: October 17, 1995.

Description of amendment request: The proposed amendment would

modify the Palisades Facility Operating License to reference 10 CFR

Part 40, allow the use of source materials as reactor fuel, delete

references to specific amendments and specific revisions in the listed

titles of the Physical Security Plan Suitability Training and

Qualification Plan and the Safeguards Contingency Plan, delete

paragraph 2.F on reporting requirements, and make minor editorial

changes. In addition, the Technical Specifications (TS) would be

modified as follows: (1) TS 3.1.2 would be modified to change the

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

shield cooling system requirements would be relocated to the Palisades

Final Safety Analysis Report (FSAR); (3) several minor editorial

changes to various sections of the TS are proposed; and (4) revisions

to several TS bases pages are proposed.

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:

Administrative Changes

Since these changes have no effect on the physical plant or its

operation, they cannot involve a significant increase in the

probability or consequences of an accident previously evaluated,

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

any previously evaluated, or involve a significant reduction in a

margin of safety.

Technical Changes

The following evaluation supports the finding that operation of

the facility in accordance with the two non-administrative changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Use of Source Material as reactor fuel: The use of depleted or

natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in

addition to the currently allowed ``slightly enriched uranium''

would not affect the physical plant or its operation in any way

which could increase the probability of any previously evaluated

accident. Its use would not introduce any new kind or additional

amount of fission product material. Therefore, use of source

material as reactor fuel would not affect the consequences of an

accident previously evaluated.

Restoration of the Pressurizer Cooldown Rate Limit: The

Palisades Technical Specifications contain a single limit, item

3.1.2 b, for both heatup and cooldown rates for the pressurizer. The

October 5, 1994 change request proposed changing that limit from

200 deg.F/hour to 100 deg.F/hour solely due to its inconsistency

with the pressurizer design analysis. Fatigue calculations in the

pressurizer design analysis assumed a heatup rate of 100 deg.F/hour

and a cooldown rate of 200 deg.F/hour. Until issuance of Amendment

163, the Technical specifications contained a single limit for both

heatup and cooldown rates of 200 deg.F/hour. Although the installed

equipment is not capable of exceeding the 100 deg.F/hour heatup

limit, the October 5, 1994 change request proposed a revised limit

to assure that the Technical Specification limit was not less

restrictive than the design analysis. The higher pressurizer

cooldown rate does not affect the results of our analyses which

determined the PCS Pressure-Temperature limits or the [Loss of

Temperature Overpressurization] LTOP setting requirements of the

Technical Specifications.

When the change was proposed, it was not realized that the more

limiting cooldown rate might adversely, and unnecessarily, affect

plant operation. This proposed change to the Technical

Specifications would separate the limits for heatup rate and

cooldown rate, returning the specified cooldown rate to the original

value which was consistent with plant design. The current heatup

rate limit, which is also consistent with the design, would be

retained. The proposed pressurizer cooldown rate will allow

depressurizing of the primary coolant system [PCS] and flooding the

pressurizer steam space without undue restriction. The more rapid

depressurization would be important in the event of a steam

generator tube rupture.

Therefore, operation of the facility in accordance with the

proposed change to the Technical Specifications would 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 previously evaluated.

Use of Source Material as reactor fuel: The use of depleted or

natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in

addition to the currently allowed ``slightly enriched uranium''

would not affect the design (other than the fuel enrichment),

configuration, or operation of the plant. Therefore this change

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

from any previously evaluated.

Restoration of the Pressurizer Cooldown Rate Limit: The proposed

change to the Technical Specifications would bring the plant within

the assumptions of the design documents for the pressurizer and in

line with the Accident analysis for the rapid reduction of the

primary coolant system pressure. With the lower rate specified in

the present technical specification, the depressurization of the PCS

will be delayed to maintain the lower pressurizer cooldown rate.

Therefore, operation of the facility in accordance with the

proposed change to the

[[Page 58400]]

Technical Specifications would not create the possibility of a new or

different kind of accident from any previously evaluated.

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

Use of Source Material as reactor fuel: The use of depleted or

natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in

addition to the currently allowed ``slightly enriched uranium''

would not affect the Safety Limits, Limiting Conditions for

Operation or other operating limits, or the safety analyses which

they support. Therefore, the margin of safety is unaffected.

Restoration of the Pressurizer Cooldown Rate Limit: The proposed

change to the Technical Specifications would bring the plant in line

with the design analysis. This will not reduce the margin of safety

since the higher rate is the basis for the present margin of safety.

Therefore, the proposed change to the Technical Specifications

would not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

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

NRC Project Director: Brian E. Holian, Acting.

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

Michigan

Date of amendment request: September 20, 1995.

Description of amendment request: The proposed amendment would

allow a one-time extension of the 18-month surveillance intervals

contained in the Technical Specifications (TS) related to system

testing, instrumentation calibration, component inspection, component

testing, response time testing and logic system functional tests for

various systems, components and 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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed TS changes involve a one-time only change in the

surveillance testing intervals to facilitate a one-time only change

in the Fermi 2 operating cycle. The proposed TS changes do not

physically impact the plant nor do they impact any design or

functional requirements of the associated systems. That is, the

proposed TS changes do not significantly degrade the performance or

increase the challenges of any safety systems assumed to function in

the accident analysis. The proposed TS changes affect only the

frequency of the surveillance requirements and do not impact the TS

surveillance requirements themselves. In addition, the proposed TS

changes do not introduce any new accident initiators since no

accidents previously evaluated have as their initiators anything

related to the change in the frequency of surveillance testing.

Also, the proposed TS changes do not significantly affect the

availability of equipment or systems required to mitigate the

consequences of an accident because of other, more frequent testing

or the availability of redundant systems or equipment. Furthermore,

a historical review of surveillance test results support the above

conclusions. Therefore, the proposed TS changes do not significantly

increase the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes involve a one-time only change in the

surveillance testing intervals to facilitate the one-time only

change in the Fermi 2 operating cycle. The propose TS changes do not

introduce any failure mechanisms of a different type than those

previously evaluated since there are no physical changes being made

to the facility. In addition, the surveillance test requirements

themselves will remain unchanged. Therefore, the proposed TS changes

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

from any previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

Although the proposed TS changes will result in an increase in

the interval between some surveillance tests, the impact, if any, on

system availability is small based on other, more frequent testing

or redundant systems or equipment, and there is no evidence of any

time dependent failures that would impact the availability of the

systems. Therefore, the assumptions in the licensing basis are not

impacted, and the proposed TS changes do not significantly reduce 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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

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

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: Brian E. Holian, Acting.

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 amendment request: August 8, 1995.

Description of amendment request: The amendments would revise

Technical Specification Section 3/4.4.8, Table 4.4-4, Table Notations,

to allow the reactor coolant system gross specific activity measurement

method to be changed from the current degassed method to a non-

degassed, or pressurized dilution, method.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The amendments will have no effect on the probability of

the occurrence of any accident. It has been demonstrated that the

results obtained by the pressurized dilution technique are

statistically similar to results obtained by the degassed technique.

Therefore, implemention of the new method will have no effect

insofar as the accuracy of the NC [reactor coolant system] system

specific activity determination is concerned. Therefore, there will

be no effect upon any accident dose consequences.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No accident causal mechanisms will be affected by

installation of the sampling equipment required by the pressurized

dilution technique. Operation of the NC system itself will not be

affected by the proposed change in sampling technique. All procedure

changes required for implementation of the new sampling method will

be made according to the provisions of 10 CFR 50.59. No impact on

other areas of plant operations will be generated as a result of the

new sampling method.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. No impact on any safety limits will

result from the change in sample method from the degassed technique

to the pressurized dilution technique. Several benefits will result

from the change,

[[Page 58401]]

including fewer opportunities for valve mispositionings to occur, as

well as reduced radiation exposure to Chemistry technicians. The

proposed amendment is consistent with a similar amendment approved

by the NRC for McGuire Nuclear Station (Amendment Nos. 66 and 47 for

McGuire Units 1 and 2, respectively).

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730.

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 7, 1995.

Description of amendment request: The proposed change would revise

Technical Specification 3/4.5.1 SAFETY INJECTION TANKS (SITs) by

increasing the specified range associated with SIT water level and

nitrogen cover pressure.

The current limiting conditions for operation (LCO) for the SIT

requires that four SITs be operable with a water volume in the range of

1679 cubic feet (78%) to 1807 cubic feet (83.8%) and a nitrogen cover

pressure between 600 psig to 625 psig. The proposed change requests an

expanded range of 925.6 cubic feet (40%) to 1807 cubic feet (83.8%) for

SIT level and 600 psig to 670 psig for SIT pressure indicators.

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 facility in accordance with this change does

not involve an increase in the probability of any accident. The SITs

are used to mitigate the consequences of an accident and are not

accident initiators.

The proposed change would actually decrease the consequence of

events such as LOCA [loss of coolant accident] which would result in

rapid RCS [reactor coolant system] depressurization.

By reducing SIT level, the initial nitrogen gas volume is

increased which results in an increase in the SIT flow rate into the

RCS for a given RCS pressure transient. This decreases the time

required to fill the reactor vessel lower plenum after the end of

blowdown. During refill, fuel cladding temperature increases rapidly

due to insufficient cooling which is provided solely by rod to rod

thermal radiation. Decreasing the refill time therefore, results in

lower cladding temperature at the start of core reflood which

results in lower Peak Cladding Temperature (PCT) during reflood.

Increasing the nitrogen cover pressure would also result in

increased SIT flow rate and would be beneficial as described above.

Therefore, the proposed change will not involve a significant

increase in the probability or consequence of any accident.

The proposed change will not create any new system connections

or interactions. Thus, no new modes of failure are introduced. The

increased range for SIT pressure and level is actually beneficial in

maintaining lower PCT following a LOCA.

Therefore, the proposed change will not create the possibility

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

evaluated.

The impact of the proposed changes on the Waterford 3 FSAR

[Final Safety Analysis Report] analyses have been evaluated. The AOR

[Analysis of Record] shows that PCT and maximum cladding oxidation

would increase slightly as a result of this change. However, they

both remain below the acceptance criteria values of 2200 degrees

fahrenhit and 17% for PCT and maximum cladding oxidation,

respectively. The system capabilities to mitigate the consequences

of accidents will be the same as they were prior to these changes.

Therefore, the proposed changes do[es] 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: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street NW, Washington, DC 20005-3502.

NRC Project Director: William D. Beckner.

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

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

Date of amendment request: August 10, 1995

Description of amendment request: This amendment would incorporate

certain improvements into the Three Mile Island, Unit 1 Technical

Specifications consistent with the Standard Technical Specifications

for Babcock and Wilcox plants. The requested changes would affect the

reactor building isolation instrumentation, sampling frequency for the

sodium hydroxide tank, and the surveillance requirements for the plant

vital bus batteries.

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:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed amendment involves changes to the

TMI-1 Technical Specifications [TS] which are consistent with the

[Babcock & Wilcox] B&W Standard Technical Specifications ([R]STS),

NUREG-1430. This change does not involve any change to system or

equipment configuration. The proposed amendment revises certain

surveillance requirements, or extends certain surveillance

intervals. The reliability of systems and components relied upon to

prevent or mitigate the consequences of accidents previously

evaluated is not degraded by the proposed changes. Therefore, this

change does not involve a significant increase in the probability of

occurrence or the 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 change

only involves changes to surveillance requirements that are

consistent with RSTS or deletion of requirements which are not

appropriate for TS. No new failure modes are created and thus the

changes are bounded by accidents previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. These proposed changes involve deletions of requirements or

changes in surveillance requirements consistent with the B&W RSTS.

No operating limits are affected and no reduction in the margin of

safety is involved.

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: Law/Government Publications

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

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

[[Page 58402]]

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

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

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: October 24, 1995.

Description of amendment request: The proposed amendment would

revise the Technical Specification (TS) Surveillance Requirement of

Section 4.4.5.1, ``Steam Generators'' and the Bases for Section 3/

4.4.5, ``Steam Generators.'' Typographical errors in Section

4.4.5.1.3.c.1 and Table 4.4-6 are also proposed to be corrected. The

proposed amendment would defer the next required surveillance to

inspect steam generator tubes from October 20, 1996, to the next

refueling outage or no later than October 20, 1997, whichever is

earlier.

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:

Pursuant to 10 CFR 50.92, NNECO [the licensee] has reviewed the

proposed one-time change to extend the maximum allowable inspection

interval for steam generator tubes from 24 months to 36 months.

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.

This change involves one-time deferment of the eddy current

inspection of the steam generator tubes until the end of the next

refueling outage following the thirteenth fuel cycle, but no longer

than 12 months beyond the original due date for the inspection. The

steam generator tubes have only been exposed to one operating cycle

and are made of thermally treated Alloy 690, one of the most

corrosion resistant material currently used in recirculating steam

generators. Following the first full fuel cycle of operation, the

steam generator tube inspection found the tubes to be in excellent

condition (i.e., no repairs were required and there was no evidence

of an active degradation mechanism). Accordingly, no significant

tube degradation is expected by the end of the thirteenth fuel

cycle. Therefore, the proposed change 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.

This one-time change, allowing the steam generator tubes to be

examined at the end of the refueling outage following Cycle 13 does

not alter the physical design, configuration, or method of operation

of the plant. The extension of the inspection interval is not

expected to result in significant steam generator tube degradation.

Therefore, the proposed change 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.

Steam generator tube degradation occurs primarily during

operation. The change to extend the maximum allowable inspection

interval for steam generator tubes from 24 months to 36 months will

not significantly increase the total operating time during Cycle 13

(the plant was in an outage for at least 10 months of the 12 month

extension). Therefore, there is no significant effect on the extent

and severity of tube degradation. The improved corrosion resistance

of the steam generators tubes (thermally treated Alloy 690)

minimizes the threat of primary- and secondary-side corrosion. No

indications of corrosion have been identified in inspections

performed so far. Based on our assessment of the inspection data and

corrosion potential, all tubes are expected to be within the

Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam

Generator Tubes,'' limits by the end of Cycle 13. Also, correction

of the typographical errors will improve the fidelity of the

specification. Therefore, this change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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. 50-423, Millstone

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

Date of amendment request: May 1, June 14 and 29, July 14, 17, 18,

and 26, 1995 with supplemental information provided by letter dated

October 20, 1995.

Description of amendment request: Each proposed amendment would

change the surveillance requirement frequency from the current once per

18-month interval to once per 24-month which is the current length of a

Millstone Unit 3 refueling cycle. The changes pertain to the following

equipment:

May 1, 1995, Flow Paths--Operating; Position Indication System; Rod

Drop Time; Seismic Monitoring System; Loose Part Detection System;

Quench Spray System; Containment Recirculation Spray System;

Containment Isolation Valves. This notice supersedes the notice

published in the Federal Register on June 6, 1995 (60 FR 29882)

relating to containment isolation valves.

May 1, 1995, Steam Generator Tube Inspections; 10CFR50, Appendix J,

Type B and Type C Tests.

June 14, 1995, AC Sources Operating; DC Sources Operating;

Containment Penetration Conductor Overcurrent Protective Devices;

Motor-Operated Valves Thermal Overload Protection.

June 29, 1995, Electric Hydrogen Recombiners; Auxiliary Feedwater

System; Reactor Plant Component Cooling Water System; Service Water

System; Snubbers.

July 14, 1995, ECCS Subsystems--Tavg Greater Than or Equal to 350

deg.F; pH Trisodium Phosphate Storage Baskets.

July 17, 1995, Supplementary Leak Collection and Release System;

Control Room Emergency Ventilation System; Control Room Envelope

Pressurization System; Auxiliary Building Filter System; Fuel Building

Exhaust Filter System.

July 18, 1995, Reactor Coolant System.

July 26, 1995; Reactor Trip System Instrumentation; ESFAS

Instrumentation; Remote Shutdown Instrumentation; Accident Monitoring

Instrumentation; RCS Total Flow Rate; Process and Radiation Monitoring

Instrumentation.

In addition, the specifications are changed from a five-column to a

one-column format.

Basis for proposed no significant hazards consideration

determination: The Commission has made a proposed determination that

the amendment request involves 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

[[Page 58403]]

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

by 10 CFR 50.91(a), the licensee has provided its analysis of the issue

of no significant hazards consideration. The NRC staff has reviewed the

licensee's analysis against the standards of 10 CFR 50.92(c). The NRC

staff's review is presented below:

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

probability or consequences of an accident previously evaluated.

The proposed changes to surveillance requirements of the Millstone

Unit No. 3 Technical Specifications extend the frequency for checking

the operability of the affected components/equipment. The proposal

would extend the frequency from at least once per 18 months to at least

once each refueling interval (i.e., nominal 24-months).

Changing the frequency of surveillance requirements from at least

once per 18 months to at least once each refueling interval does not

change the basis for the frequency. The frequency was chosen because of

the need to perform this verification under the conditions that apply

during a plant outage, and to avoid the potential of an unplanned

transient if the surveillances were conducted with the plant at power.

The proposed changes do not alter the intent or method by which the

surveillances are conducted, do not involve any physical changes to the

plant, do not alter the way any structure, system, or component

functions, and do not modify the manner in which the plant is operated.

As such, the proposed changes in the frequency of surveillance

requirements will not degrade the ability of the equipment/components

to perform its safety function.

Additional assurance of the operability of the components/equipment

is provided by additional surveillance requirements (e.g., monthly or

quarterly surveillances).

Equipment performance over the last four operating cycles was

evaluated to determine the impact of extending the frequency of

surveillance requirements. This evaluation included a review of

surveillance results, preventive maintenance records, and the frequency

and type of corrective maintenance. It concluded that there is no

indication that the proposed extension could cause deterioration in the

condition or performance of any of the subject components.

In addition to the substantive changes, there are format changes

which are merely editorial and because format changes produce no

physical change they do not influence the probability or consequences

of accidents.

Since the proposed changes only affect the surveillance frequency

for safety systems that are used to mitigate accidents, the changes

cannot affect the probability of any previously analyzed accident.

While the proposed changes can lengthen the intervals between

surveillances, the increases in intervals has been evaluated and it is

concluded that there is no significant impact on the reliability or

availability of the safety system and consequently, there is no impact

on the consequences on any analyzed accident.

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

kind of accident from any accident previously evaluated.

The proposed changes to surveillance requirements of the Millstone

Unit No. 3 Technical Specifications extend the frequency for verifying

the operability of the affected components/equipment. The proposal

would extend the frequency from at least once per 18 months to at least

once each refueling interval (nominal 24 months).