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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from September 2, 1994, through September 16,

1994. The last biweekly notice was published on September 14, 1994 (59

FR 47163).

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

20555. The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By October 28, 1994, 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 20555 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 internvetion 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 Document Room,

the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 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)(l)(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 20555, and at the local public document

room for the particular facility 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 amendment request: August 25, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification Table 2.2-1, Reactor Trip System

Instrumentation Trip Setpoints, and Table 3.3-4, Engineered Safety

Actuation System Instrumentation Trip Setpoints, to reflect a revised

steam generator [water] level process measurement accuracy. The steam

generator level trip setpoints are not affected by the proposed

amendment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The Technical Specification Tables 2.2-1 and 3.3-4 which

document Total Allowable, Z, and S values are being revised to

reflect additional Process Measurement Error uncertainties based

upon enhanced knowledge of steam generator performance provided by

the Nuclear Steam Supply System (NSSS) vendor. There will be no

physical changes to plant equipment, logic, or control as a result

of the proposed amendment. The safety-related trip setpoints are not

being changed. Therefore, there would be no increase in the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The propose amendment does not introduce any new equipment,

logic, or control functions. The steam generator water level

protective setpoints are not being changed. No new common mode

failure mechanism is being introduced. Therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed amendment revises Technical Specification Tables

2.2-1 and 3.3-4 to more accurately reflect PMA uncertainties based

upon enhanced knowledge of the steam generator performance provided

by the NSSS vendor. The margin of safety as defined in the Technical

Specifications is not reduced by the proposed changes to the Tables.

Calculations demonstrate that this requirement is still satisfied

with the new values based upon the enhanced understanding of PMA

terms. 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: Cameron Village Regional

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

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

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

NRC Project Director: David B. Matthews.

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

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

Date of amendment request: August 25, 1994.

Description of amendment request: The requested amendments allow

the testing interval for auxiliary feedwater (AFW) system pumps to be

increased from monthly to quarterly on a staggered test basis. The

proposed amendments are consistent with NRC staff recommendations and

guidance contained in NUREG-1366, ``Improvements to Technical

Specifications Surveillance Requirements'' and Generic Letter 93-05,

``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation.''

In addition, the requested amendments incorporate a note from

Surveillance Requirement 3.7.5.2 of NUREG-1431, ``Revised Standard

Technical Specifications, Westinghouse Plants'' into the existing

McGuire Technical Specifications governing AFW system pump testing.

This note clarifies that the turbine-driven AFW pump cannot be tested

until the required pressure exists in the secondary side of the steam

generator.

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:

As required by 10 CFR 50.91, this analysis is provided

concerning whether the requested amendments involve significant

hazards considerations, as defined by 10 CFR 50.92. Standards for

determination that an amendment request involves no significant

hazards considerations are if operation of the facility in

accordance with the requested 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 requested amendments decrease from monthly to quarterly the

frequency at which the motor-driven and turbine-driven AFW pumps

must be demonstrated operable as specified in TS... 4.7.1.2

(McGuire). They also incorporate a note of clarification from the

new Westinghouse STS [Standard Technical Specifications] into the

existing... McGuire specification concerning when the pump head or

discharge pressure versus flow verification for the turbine-driven

pump is required to be performed.

In 48 FR 14870, the Commission has set forth examples of

amendments that are considered not likely to involve significant

hazards considerations. Example vii describes a change to make a

license conform to changes in regulations, where the license change

results in very minor changes to facility operations clearly in

keeping with the regulations. The requested amendments are similar

to example vii in that they result in minor changes to plant

surveillance requirements and are consistent with the existing NRC

position and guidance contained in NUREG-1366 and Generic Letter 93-

05, as well as NUREG-1431. While the issuance of NUREG-1366 and

Generic Letter 93-05, as well as NUREG-1341 does not constitute a

change in existing regulations, it nevertheless establishes the NRC

staff's position concerning the acceptability of decreasing the

surveillance frequency of AFW pumps from monthly to quarterly and

concerning the acceptability of adopting all or part of the new STS.

The requested amendments are consistent with the position of NUREG-

1366 and with the guidance of Generic Letter 93-05, as well as with

NUREG-1431.

Criterion 1

The requested amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Decreasing the frequency of AFW pump testing as specified

in TS from monthly to quarterly will have no impact upon the

probability of any accident, since the AFW pumps are not accident

initiating equipment.

Also, since... McGuire's AFW pump performance histories support

making the proposed change, system response following an accident

will not be adversely affected. Therefore, the requested amendments

will not result in increased accident consequences. ***Incorporating

the new STS note will only serve to clarify when the turbine-driven

pump is required to be tested and will not have any impact upon

either the probability or consequences of any accident. The pump

will still be tested as before and its acceptance criteria will be

unaffected.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated above, the AFW pumps are not accident

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

accident standpoint. The plant will not be operated in different

manner. ***Incorporating the clarifying note from the new STS will

not result in any new accident sequences, since plant operation will

be unaffected.

Criterion 3

The requested amendments will not involve a significant

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

unaffected by the proposed changes. The AFW pumps will still be

capable of fulfilling their required safety function, since plant

operating experience supports the proposed change. The availability

of the AFW pumps will be increased as a result of the proposed

amendments because they will not have to be made unavailable for

testing as frequently. Finally, the proposed amendments are

consistent with the NRC position and guidance set forth in NUREG-

1366 and Generic Letter 93-05. ***Incorporating the note from the

new STS will not impact any safety margins.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

The proposed technical specification amendment has been reviewed

against the criteria of 10 CFR 51.22 for environmental

considerations. The proposed amendment does not involve a

significant hazards consideration, nor increase the types and

amounts of effluents that may be released offsite, nor increase

individual or cumulative occupational radiation exposures.

Therefore, the proposed amendment meets the criteria given in 10 CFR

51.22(c)(9) for a categorical exclusion from the requirement for an

Environmental Impact Statement.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

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

Date of amendment request: June 13, 1994.

Description of amendment request: The proposed amendment would

modify Clinton Power Station Technical Specification 3/4.6.1.8,

``Containment Building Ventilation and Purge Systems,'' which includes

a requirement to perform a leak rate measurement at least once per 92

days on each 36-inch supply and exhaust containment ventilation

isolation valve with a resilient seal. The proposed modification would

require a leak test at least once per 18 months provided the valves

remain closed during that period.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

(1) The proposed change does not involve a change in plant

design. Failure of or leakage through a containment barrier cannot

itself create an accident; therefore, this change would not increase

the probability of any accident previously evaluated. Failure of or

leakage through a containment barrier can, however, increase the

consequences of those accidents previously evaluated. The proposed

change merely revises the frequency at which the local leak rate

test is performed on the containment building HVAC 36-inch supply

and exhaust penetrations. The containment isolation valves for these

penetrations are normally only opened during refueling outages. The

stroke testing for the isolation valves has been changed to a cold

shutdown frequency and, as a result, there is no mechanism present

to degrade the seals and cause increased leakage through the

penetration. Based on past penetration leak rate measurements, it

has been determined that leak rate testing on an 18-month frequency

is sufficient to identify seal degradation if the valves are not

opened. However, should the valves be opened during the 18-month

interval, the proposed change would require that a leak rate test be

performed within 92 days. This will ensure that the leak rate for

the given penetration has not exceeded the specified limit as a

result of stroking the valve. Penetration leakage will continue to

be measured at sufficient intervals to identify seal degradation in

the 36-inch containment isolation valves. In addition, the same

leakage limits will be imposed. Therefore, the proposed change will

not result in a significant increase in the probability or the

consequences of any accident previously evaluated.

(2) This request does not result in any change to the plant

design nor does it involve a change in current plant operation. The

proposed change will not change the design basis for the valves

being leak tested. The valves will continue to be verified to meet

the required leak rate and the safety function of the subject valves

remains unchanged. Furthermore, any potential leakage through the

containment building HVAC 36-inch supply exhaust and supply

penetrations cannot create an accident. As a result, the proposed

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

accident from any accident previously evaluated.

(3) The only margin of safety that could potentially be impacted

by the proposed change to the surveillance requirement frequency is

the margin concerning the offsite dose consequences of postulated

accidents (which is directly related to the containment leak rate).

As discussed above, this request does not result in a significant

increase in the consequences of any accident previously evaluated.

It has been demonstrated that the penetration leakage does not

change appreciably when the valves are not stroked. Therefore, since

the valves are normally only opened during refueling outages,

leakage through the penetrations is not expected to change during

the proposed 18-month interval between leak rate tests. Should the

valves be opened during the 18-month interval, a local leak rate

test will be performed within 92 days. The proposed leak rate test

frequency will provide sufficient indication of seal degradation to

allow the opportunity for repair before gross leakage failures

develop. In addition, the proposed change involves no change to the

currently established leak rate test acceptance criteria. As a

result, the proposed changes do not result in 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th St., Decatur,

IL 62525.

NRC Project Director: John N. Hannon.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

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

Date of amendment request: August 12, 1994.

Description of amendment request: The proposed amendment would

modify Clinton Power Station Technical Specification 3/4.6.2.2,

``Drywell Bypass Leakage,'' to allow drywell by pass leakage rate tests

(DBLRTs) to be performed at intervals as long as ten years based on the

demonstrated performance of the drywell structure. DBLRTs are currently

required to be performed once every 18 months.

Basis of 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 change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that contribute to

initiation of any accidents previously evaluated. Thus, the proposed

change cannot increase the probability of any accident previously

evaluated.

The proposed change potentially affects the leak tight integrity

of the drywell, a structure used to mitigate the consequences of a

loss of coolant accident (LOCA). The function of the drywell is to

channel the steam released from the LOCA through the suppression

pool, limiting the amount of steam released to the primary

containment atmosphere. This limits the containment pressurization

due to the LOCA. The leakage of the drywell is limited to ensure

that the primary containment does not exceed its design limits of

185 deg.F and 15 psig. Because the proposed change does not alter

the plant design, only the frequency of measuring the drywell

leakage, the proposed change does not directly result in an increase

in the drywell leakage. However, decreasing the test frequency can

increase the probability that a large increase in drywell bypass

leakage could go undetected for an extended period of time. There

are several potential sources of steam bypass leakage paths. These

include potential cracks in drywell concrete structure, the drywell

vacuum breakers, and various penetrations through the drywell

structure. Based on the results of the structural integrity test

conducted as part of the preoperational test program, additional

cracking of the drywell is not expected during the remaining life of

the plant. Ventilation and piping penetrations (including the

drywell vacuum breaker penetrations) are designed to ASME Code Class

2 and Seismic Category 1 requirements. These penetrations are

designed with two isolation valves in series with one valve in the

drywell and another either outside primary containment or in the

wetwell. High energy lines that extend into the wetwell, such as the

main steam lines and feedwater lines, are encapsulated by guard

pipes to direct energy to the drywell in case of a piping rupture.

Electrical penetrations are sealed with a high strength/density

material that will prevent leakage as well as provide radiation

shielding. Operational experience has shown that the leak tightness

of the drywell has maintained well below the allowable leakage

limits. In fact, the calculated drywell bypass leakage area is of

such a small magnitude that containment design pressure could not be

exceeded even if containment spray and heat sinks were not

available. The technical specification limit of 10% of the maximum

allowable leakage path area provides margin for degradation. Drywell

performance data to date suggest that drywell degradation, even

during a ten-year interval between tests, will not exceed this

margin.

Further, an analysis was conducted to determine the potential

risk to the public from unacceptable drywell bypass leakage going

undetected as a result of the proposed change. Based on this

analysis, under several different accident scenarios, the risk of

radioactivity release from containment was found to be negligible,

about 10-9 per year.

Based on the above, Illinois Power has concluded that the

proposed change will not result in a significant increase in the

consequences of any accident previously evaluated.

(2) The proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that could contribute to

initiation of any accidents. Thus, the proposed change cannot create

the possibility of an accident not previously evaluated.

(3) The proposed change only affects the frequency of measuring

the drywell leakage and does not change the bypass leakage limit for

the drywell. However, the proposed change can increase the

probability that a large increase in drywell bypass leakage could go

undetected for an extended period of time. Operational experience

has shown that the leak tightness of the drywell has been maintained

well below the allowable leakage limits. In fact, the calculated

drywell bypass leakage area is of such a small magnitude that

containment design pressure could not be exceeded even if

containment spray and heat sinks were not available. Further, an

analysis was conducted to determine the potential risk to the public

from the proposed change. Based on this analysis, under several

different accident scenarios, the risk of radioactivity release from

containment was found to be negligible, about 10-9 per year. As

a result, Illinois Power has concluded that the proposed change will

not result in 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th St., Decatur,

IL 62525.

NRC Project Director: John N. Hannon.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

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

Date of amendment request: August 12, 1994.

Description of amendment request: The proposed amendment would

modify Clinton Power Station Technical Specifications 3/4.3.1,

``Reactor Protection System Instrumentation,'' 3/4.3.2, ``Containment

and Reactor Vessel Isolation Control System,'' 3/4.3.3, ``Emergency

Core Cooling System Actuation Instrumentation,'' 3/4.3.4.2, ``End-of-

Cycle Recirculation Pump Trip System Instrumentation,'' 3/4.3.5,

``Reactor Core Isolation Cooling System Actuation Instrumentation,'' 3/

4.4.2.1, ``Safety/Relief Valves,'' and 3/4.4.2.2, ``Safety/Relief

Valves Low-Low Set Function.'' These technical specifications contain

requirements to perform manual testing of the associated solid-state

logic at least once every four fuel cycles. This testing is in addition

to the automatic testing performed by the self-test system. Due to the

negative impact on plant safety caused by the need to remove systems

from service to prevent unwanted actuations and the increased potential

for unintended equipment actuation during manual testing, Illinois

Power is proposing that the requirement to perform manual testing of

the solid-state logic independently from the self-test system be

eliminated.

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 change to the plant

design. The proposed change involves only testing of the solid-state

Nuclear Systems Protection System (NSPS) utilizing the self-test

system (STS). As identified in Supplement No. 2 to the Clinton Power

Station (CPS) Safety Evaluation Report (SSER 2), use of the STS to

perform certain surveillance testing required by the plant Technical

Specifications is acceptable. However, as noted in SSER 6, portions

of the NSPS logic must also be manually tested during each refueling

outage independently from the STS such that all NSPS trip/actuation

functions are tested independently of the STS at least once every

four fuel cycles. The change proposed in this request consists of

the elimination of this manual testing of the NSPS logic

independently from the STS.

As identified in SSER 6, the purpose of the currently required

manual tests is to provide a means to verify operability of the NSPS

functional circuits independent of the STS, and thereby (1) detect

any failures undetected by the STS and take corrective action to

restore proper operation of the NSPS and STS and (2) assuming that

no additional failures beyond those identified by the STS are

detected during independent logic testing, confirm the validity of

the STS test results. The ability of the STS to detect functional

failures of the NSPS logic as designed was verified as part of

General Electric's independent design verification. In addition,

this capability was also verified by testing as part of the

preoperational test program as described in Section 14.2.12.1.62 of

the CPS Updated Safety Analysis Report (USAR) and during testing

performed during the first four refueling outages in accordance with

the current Technical Specification testing requirements. Thus,

Illinois Power (IP) has concluded that all functional failures

undetectable by the STS have been identified and the aforementioned

objectives have been satisfied.

Since the proposed change does not alter the plant design or

operation, it cannot increase the probability of any accident

previously evaluated. This proposed change does involve the NSPS

logic which is utilized to actuate systems needed to mitigate the

consequences of accidents previously evaluated, however, the

proposed change merely eliminates the currently required manual

testing independently from the STS once every four fuel cycles.

Since the proposed change does not alter the NSPS logic, does not

impact operation of the STS, and continues to require adequate

testing of the NSPS logic on a frequency sufficient to maintain the

operability of the associated NSPS logic, the proposed change cannot

impact the reliability of the associated actuation instrumentation

and therefore cannot increase the consequences of any accident

previously evaluated.

(2) Adequate testing of the NSPS logic will continue to be

required. The proposed change continues to allow us of the STS in

performing surveillance tests as documented by the NRC in SSER 2.

However, additional manual tests independently from the STS will no

longer be required. Since the proposed change does not add

additional testing configurations or operating modes nor does it

alter the plant design, it will not introduce any new failure modes.

Thus, this proposed change cannot create the possibility of a new or

different kind of accident from any accident previously evaluated.

(3) Performance of the currently required manual testing of the

NSPS logic independently from the STS system involves actuation

logic for the reactor protection system, emergency core cooling

systems (ECCS), reactor core isolation cooling system, automatic

depressurization system, nuclear steam supply shutoff system, and

the residual heat removal (RHR) system, including the shutdown

cooling mode of operation. Performance of these tests requires these

systems to be disabled to prevent unwanted system actuations.Thus,

these systems are rendered inoperable during this testing. This also

results in extensive temporary reconfiguration of systems and

actuation instrumentation, including logic card removal,

installation of signal simulators, disconnecting load drivers, etc.

The removal of these safety systems from service to perform this

testing results in reduced availability of RPS, ECCS, and RHR

shutdown cooling systems during the plant outage. In addition to the

intentional disabling of equipment to perform this testing, this

testing has in the past led to safety system unavailability due to

equipment damage caused by bending pin connectors and burning

circuit cards out due to electrical shorting. Further, CPS has

experienced unintentional equipment actuations resulting in

unnecessary challenges to safety systems and the need to file

licensee event reports with the NRC. IP has concluded that

elimination of this currently required manual testing of the NSPS

will have a positive impact on overall plant safety.

The only margin of safety that could be negatively impacted by

this proposed change is the potential for a functional failure in

the NSPS logic going undetected. The manual tests proposed for

deletion are only required to be performed at least once every four

fuel cycles. In addition, the ability of the STS to detect

functional failures of the NSPS logic as designed was verified as

part of General Electric's independent design verification. Further,

this capability was also verified as part of the preoperational test

program as described in Section 14.2.12.1.62 of the CPS Updated

Safety Analysis Report (USAR) and subsequently during testing

performing during the first four refueling outages. These manual

tests did not identify any functional failures of the NSPS logic

which would be expected to be detected by the STS per its design.

All other functional failures undetectable by the STS have been

identified as Untested Islands (UTIs). Thus, the original objectives

of this testing have been satisfied.

The proposed change will continue to require testing at

refueling outage intervals. As identified in USAR Section

7.2.1.1.4.8 and SSER 6, circuits which are not capable of being

tested by the STS are identified as UTIs. General Electric and IP

have identified, via analysis and manual testing, all UTIs in the

functional NSPS logic and have established procedures for testing

these UTIs. Periodic verification of the operability of these UTIs

will continue to be performed at the frequencies recommended by the

manufacturer as accepted by the NRC in SSER 6. These frequencies

were established by the manufacturer based on mean time between

failure analyses for the components in the associated circuits and

may be as long as six years. Based on the above, the functional

operability of the NSPS logic is adequately assured.

From the above, IP has concluded that the proposed change will

result in a net increase in the overall 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th St., Decatur,

IL 62525.

NRC Project Director: John N. Hannon.

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

D. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

Date of amendment requests: April 6, 1994.

Description of amendment requests: The proposed amendments would

remove the license condition reference to Table 1 of the Fire

Protection Safety Evaluation Report (SER) for Cook Nuclear Plant,

approved on June 4, 1979, and issued under a cover letter dated July

31, 1979. Table 1 of the 1979 First Protection SER is a schedule for

completion of 23 modifications which have since been completed. Three

of the modifications (Nos. 7C, 9, & 20) have been changed since the

1979 Fire Protection SER.

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

This update of the fire protection modifications contained in

Table 1 of the July 31, 1979, fire protection SER does not involve a

significant increase in the probability or consequences of an

accident previously analyzed as follows:

Item 7C--The use of an air compressor rather than the cascade

recharging stations does not adversely impact the ability to supply

breathing air for the fire brigade and is acceptable under BTP APCSB

9.5-1.

Item 9--the use of unrated metal hatches has been accepted by

the NRC in a SER date June 17, 1988.

Item 20--The new fire pumps have already been accepted by the

NRC in the SER dated March 31, 1993, Amendment Nos. 171 and 154. The

retirement of the old screenhouse diesel driven fire pumps is based

on the availability of another supply of ``backup'' fire suppression

water.

Criterion 2

This update of the fire protection modifications contained in

Table 1 of the July 31, 1979, fire protection SER does not create

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

accident previously evaulated as follows:

Item 7C--The use of the an air compressor rather than the

cascade recharging stations does not adversely impact the ability to

supply breathing air for the fire brigade and is acceptable under

BTP APCSB 9.5-1.

Item 9--The use of unrated metal hatches has been accepted by

the NRC in a SER dated June 17, 1988.

Item 20--The new fire pumps have already been accepted by the

NRC in the SER dated March 31, 1993, Amendment Nos. 171 and 154. The

retirement of the old screenhouse diesel driven fire pumps is based

on the availability of another supply of ``backup'' fire suppression

water.

Criterion 3

This update of the fire protection modifications contained in

Table 1 of the July 31, 1979, fire protection SER does not involve a

significant reduction in a margin of safety as follows:

Item 7C--The use of an air compressor rather than the cascade

recharging stations does not adversely impact the ability to supply

breathing air for the fire brigade and is acceptable under BTP APCSB

9.5-1. Filled spare breathing air bottles are on-site and the local

municipal department will provide assistance as needed.

Item 9--The use of unrated metal hatches has been accepted by

the NRC in a SER dated June 17, 1988.

Item 20--The new fire pumps have already been accepted by the

NRC in the SER dated March 31, 1993, Amendment Nos. 171 and 154. The

retirement of the old screenhouse diesel driven fire pumps is based

on the availability of another supply of ``backup'' fire suppression

water.

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: Ledyard B. Marsh

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 request: July 19, 1994

Description of amendment request: The proposed amendments would

revise the technical specifications by removing the specific scheduling

requirements for Types A, B, and C tests and replacing these

requirements with a requirement to perform Types A, B, and C testing in

accordance with Appendix J to 10 CFR 50.

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

This amendment request does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the proposed changes to the T/Ss do not affect the

assumptions, parameters, or results of any UFSAR accident analysis.

The proposed changes do not modify the response of the containment

during a design basis accident. The proposed amendment does not add

or modify any existing equipment. The proposed Types A, B, and C

testing schedules will be consistent with Appendix J to 10 CFR 50.

Based on these considerations, it is concluded that the changes do

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Criterion 2

The proposed changes do not involve physical changes to the

plant or changes in plant operating configuration. The proposed

changes only remove the restrictive schedular requirements for

conducting Type A testing from the T/Ss and substitute the schedule

specified in Appendix J to 10 CFR 50. For Types B and C testing, the

schedular requirements are removed from T/Ss because they are

already specified in Appendix J to 10 CFR 50. Thus, it is concluded

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

different kind of accident from any accident previously evaluated.

Criterion 3

The margin for safety presently provided is not reduced by the

proposed change in the schedular requirements for Type A tests.

Types B and C schedular requirements are not changed by removing

them from T/Ss. Although the changes allow more flexibility in

scheduling Type A tests, the proposed amendment continues to ensure

reactor containment system reliability by periodic testing in full

compliance with 10 CFR 50, Appendix J. Based on these

considerations, it is concluded that the 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: 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: Ledyard B. Marsh.

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 request: July 26, 1994.

Description of amendment request: The proposed amendments would

modify the technical specifications such that the requirement to

measure the moderator temperature coefficient near the end of the cycle

will become conditional. The test will not be performed if specified

core performance benchmark criteria are met for the operating cycle,

and the revised predicted moderator temperature coefficient is less

negative than the moderator temperature coefficient surveillance limit

presented in the Core Operation Limits Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

Does the EOL MTC measurement conditional exemption involve a

significant increase in the probability or consequences of an

accident previously evaluated?

No. The conditional exemption of the most negative MTC

measurement does not change the most negative MTC surveillance and

LCO limits in the T/Ss. Since these MTC values are unchanged, and

since the basis for the derivation of these values from the safety

analysis MDC is unchanged, the constant MDC assumed for the FSAR

safety analyses will also remain unchanged. Therefore, no change in

the modeling (i.e., probabilities) of the accident analysis

conditions or response is necessary in order to implement the change

to the conditional exemption methodology. In addition, since the

constant MDC assumed in the safety analyses is not changed by the

conditional exemption of the most negative MTC surveillance

measurement, the consequences of an accident previously evaluated in

the FSAR are not increased. The dose predictions presented in the

FSAR for a SGTR remain valid such that more severe consequences will

not occur. Additionally, since mass and energy releases for LOCA and

steamline break are not increased as a result of the unchanged MDC,

the dose predications for these events presented in the FSAR also

remain bounding.

Criterion 2

Does the EOL MTC measurement conditional exemption create the

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

previously evaluated?

No. Since the EOL MTC is not changed by the conditional

exemption methodology of Reference 1, the possibility of an accident

which is different than any already evaluated in the FSAR has not

been created. No new or different failure modes have been defined

for any system or component nor has any new limiting single failure

been identified. Conservative assumptions for MDC have already been

modeled in the FSAR analyses. These assumptions will remain valid

since the conditional exemption methodology documented in Reference

1 does not change the safety analysis MDC nor the T/S values of the

MTC.

Criterion 3

Does the EOL MTC measurement conditional exemption involve a

significant reduction in a margin of safety?

No. The evaluation of the conditional exemption methodology

documented in Reference 1 has taken into account the applicable Cook

Nuclear Plant Units 1 and 2 T/Ss and has bounded the conditions

under which the specifications permit operation. The applicable T/Ss

are surveillance 4.1.1.4, and reference ``e'' is added to the list

of references in Specification 6.9.1.11.2. An additional

specification 6.9.1.12 is added to define the requirements for the

``Most Negative Moderator Temperature Coefficient Limit Report,''

which is described in Appendices A, C, and D of Reference 1. The

COLR has also been modified as described in Appendix B of Reference

1. The analyses which support these T/Ss have been evaluated. The

results, as presented in the FSAR, remain bounding since the MDC

assumed in the safety analyses and the LCO and surveillance

requirement MTCs in the T/Ss remain unchanged. Therefore, the margin

of safety, as defined in the bases to these T/Ss, is not reduced.

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: 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: Ledyard B. Marsh.

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

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

Date of amendment request: August 26, 1994

Description of amendment request: The proposed amendment to

Technical Specification 4.3.3.c(1) would allow a one-time extension of

the 10 year service period for the Primary Containment Integrated

Leakage Rate (Type A) Test. Specifically, the proposed one-time only

change would extend the 10-year service interval requirement for

performance of the Type A test to correspond with the end of the

current inservice inspection interval (ISI). The interval extension

would avoid the necessity of performing an additional Type A test only

22 months after the previous test. This would result in an extension of

the second interval from 10 years to approximately 14 years and 46

months between the second and third test.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed extension of the Type A test 10-year service

interval does not increase the chances for a previously analyzed

accident to occur. Containment integrity is required for the

mitigation of accident consequences. Furthermore, containment

leakage is not the precursor to any analyzed event. Extension of the

Type A test surveillance interval will not affect the containment's

ability to maintain leakage below that assumed in the safety

analysis. The previous Type A test was completed successfully and no

plant modifications have been made or are planned (other than those

that require Type B or C testing) since the last test which could

directly affect the test results. Type B and C testing of individual

penetrations has been satisfactory and will continue to be performed

in accordance with the Technical Specifications. There have been no

pressure or temperature excursions in the containment which could

have adversely affected containment integrity. Hence, the ability of

containment to maintain leakage within the Type A test limits be

maintained.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

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

different kind of accident from any accident previously evaluated.

The proposed one-time extension of the Type A test 10-year

service interval will not affect the test methodology or acceptance

criteria nor does it alter the physical containment structure or

boundary in any way. There will be no addition or removal of plant

hardware. No new plant operating modes are being introduced. Results

of the previous Type A tests are well below allowable limits, and

there have been no plant modifications (other than those that

require Type B or C testing) since the last test nor are any

planned, that could directly impact the previous Type A test

results.

Therefore, the proposed change will not create the possibility

of a new or different accident from any previously evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

Safety margins are established through the Nine Mile Point Unit

1 safety analyses as reflected in the Technical Specification

Limiting Conditions for Operation. Containment leak rates assumed in

the safety analyses are not increased by the proposed change to the

Type A test 10-year service interval. The acceptance criteria which

must be met to verify that leak rates remain within assumed values

will not be changed.

Although the test frequency will be relaxed for the one-time

extension, no plant modifications have been made or are planned

which would invalidate the last Type A leak test results which

confirm acceptable containment integrity. Furthermore, Type B and C

testing of individual penetrations has been satisfactory and will

continue to be performed in accordance with the Technical

Specifications to assure that containment integrity is maintained.

Therefore, the proposed change will not involve a significant

reduction in a margin of a 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: Mark J. Wetterhahn, Esquire, Winston &

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

NRC Project Director: Michael J. Case.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: August 26, 1994.

Description of amendment request: The proposed change would revise

Technical Specification (TS) Section 3/4.6.1.3, ``Primary Containment

Air Locks.'' Specifically, TS 3/4.6.1.3 would be revised to allow

continued plant operation if an interlock becomes inoperable as long as

an operable door is locked shut and periodically checked as being

locked shut.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The primary containment and containment air locks are not

initiators or precursors to an accident. Primary containment

integrity ensures that the release of radioactive materials from the

containment will be restricted to those leakage paths and associated

leak rates assumed in the accident analyses. Therefore, the proposed

changes to the air lock ACTION statements cannot affect the

probability of a previously evaluated accident.

The purpose of a primary containment air lock interlock is to

allow only one door to be opened at a time in each penetration. This

provision ensures that a gross breach of primary containment does

not exist when primary containment is required to be operable.

Closure of a single door in each air lock is sufficient to provide a

leak tight barrier following postulated events. If an air lock

interlock is inoperable, the proposed ACTION requires that an

operable door be locked shut and periodically verified locked and

shut. This assures that at least one air lock door is closed, which

provides the function of the interlock, thereby assuring containment

integrity is maintained. Therefore, the proposed change will not

significantly increase the consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

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

different kind of accident from any accident previously evaluated.

The proposed changes to the primary containment air lock

specification will allow contained plant operation with an

inoperable interlock as long as an operable door is locked closed

and periodically verified to be locked closed. The changes do not

introduce any new accident precursors and do not involve any

alterations to plant configurations which could initiate a new or

different kind of accident. The change provides an alternate means

of ensuring that only one primary containment air lock door is

opened at a time in each penetration. Therefore, the proposed

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

of accident from any previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed changes to the primary containment air lock ACTION

statements will not affect the ability of the containment to respond

to an accident and limit releases to within 10 CFR Part 100 and GDC

[General Design Criterion] 19 guidelines. The changes do not affect

the design or performance characteristics of the containment or

containment air locks but simply provide an alternative means of

ensuring that only one air lock door is opened at a time. Therefore,

the proposed changes do not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Michael J. Case.

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

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

Date of amendments request: June 9, 1994.

Description of amendments request: The proposed change would revise

Section 3.4, ``Response Force Capabilities,'' of the Nine Mile Point

Nuclear Station Physical Security Plan with regard to the number of

armed Security Force Members comprising the Response Force for each

shift.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Units 1 and 2 in accordance

with the proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment to the Physical Security Plan involves a

change in the armed response force size for Units 1 and 2. Nuclear

Security's ability to meet the response requirements of 10 CFR

[subsection] 73.55(h) will not be significantly affected as was

determined utilizing NUREG-0907 while taking into consideration

several Security Program enhancements implemented since 1986 when

response force size was last reviewed.

The change does not affect the design, function, operation,

maintenance or testing of structures, systems and components at Unit

1 or Unit 2. It does not affect safety analysis or any Technical

Specification that preserves a safety analysis assumption.

Therefore, the proposed amendment will neither increase the

probability nor the consequences of an accident previously

evaluated.

The operation of Nine Mile Point Units 1 and 2 in accordance

with the proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed amendment to the Physical Security Plan involves a

change in the size of the response force for Units 1 and 2. Design,

function, operation, maintenance and testing of structures, systems

and components at Units 1 and 2 are not affected. Also, as

demonstrated using NUREG-0907, Nuclear Security will continue to

meet the requirements of 10 CFR [subsection] 73.55(h). Therefore,

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

different kind of accident from any accident previously evaluated.

The operation of Nine Mile Point Units 1 and 2 in accordance

with the proposed amendment will not involve a significant reduction

in a margin of safety.

The proposed amendment to the Physical Security Plan involves a

change in the size of the armed response force for Units 1 and 2.

The proposed change has no impact on the physical design of the

plants nor on the function or operation of their structures, systems

and components. The proposed amendment does not impact and therefore

does not reduce the margin of safety as defined in the basis for any

Unit 1 or Unit 2 Technical Specification.

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Michael J. Case.

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

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: September 1, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specifications concerning the Reactor Coolant System

Volume (RCS).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

* * * The proposed change does not involve an SHC because the

change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to Section 5.4.2 of the Millstone Unit No. 2

Technical Specifications revises the total RCS water and steam

volume to reflect the installation of new steam generators. Section

5.0 of the Millstone Unit No. 2 Technical Specifications delineates

design features for Millstone Unit No. 2. The proposed change does

not modify a limiting condition for operation, action statement, or

surveillance requirement. Additionally, the proposed change does not

revise the manner in which the plant is operated. It simply revises

the value delineated for the total RCS water and steam volume.

The steam generator replacement modifications were addressed in

a plant design change request performed in accordance with the

requirements of 10 CFR 50.59. NNECO concluded that the replacement

of the steam generators did not involve an unreviewed safety

question. This proposed change is a result of those modifications.

Based on the above, the proposed change to Section 5.4.2 of the

Millstone Unit No. 2 Technical Specifications does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any previously analyzed.

The proposed change to Section 5.4.2 of the Millstone Unit No. 2

Technical Specifications simply revises the total RCS water and

steam volume to reflect the installation of new steam generators. It

does not modify the manner in which any plant equipment or systems

are operated. Thus, the proposed change to Section 5.4.2 of the

Millstone Unit No. 2 Technical Specifications 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 Section 5.4.2 of the Millstone Unit No. 2

Technical Specifications revises the total RCS water and steam

volume to reflect the installation of new steam generators. This

proposal simply reflects the slight change in volume due to the

replacement of the steam generators. The steam generator replacement

modifications were addressed in a plant design change request

performed in accordance with the requirements of 10 CFR 50.59. NNECO

concluded that the replacement of the steam generators did not

involve an unreviewed safety question.

Based on the above, the proposed change to Section 5.4.2 of the

Millstone Unit No. 2 Technical Specifications 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 Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: John F. Stolz

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

Susquehanna Steam Electric Station, Unit 1, Luzerne County,

Pennsylvania

Date of amendment request: August 22, 1994.

Description of amendment request: The proposed amendment would

change Technical Specifications 3/4.1.3: to extend the scram discharge

volume (SDV) vent or drain valve restoration time from the current time

period of 24 hours to 7 days; to permit the SDV vent and drain valves

operability check to be performed at shutdown conditions instead of at

least once per 18 months; and to delete the SDV float switch response

surveillance.

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.

The proposed change to Technical Specification 3.1.3.1.d, which

extends the restoration time for one inoperable vent and/or drain

valve from 24 hours to 7 days, does not significantly increase the

probability or consequences of an accident previously evaluated. A 7

day restoration time, as identified in NUREG 1433, is reasonable

given the level of redundancy in the lines and the low probability

of a scram occurring while the valve is inoperable and the line is

not isolated. If the inoperable valve fails open, the redundant

valve in the line allows for leakage from the CRD [control rod

driveline] to be drained out and also allows for the line to be

isolated if necessary. If the valve fails closed, the line becomes

isolated. However, float switches and pressure sensors will notify

operators of water buildup in the instrument volume. A review of the

surveillance data indicates the vent and drain valves rarely fail

the initial operability test and require rework. The low failure

rate combined with the redundancy of the valves makes for a highly

reliable system. Therefore, the proposed change does not

significantly increase the probability or consequences of an

accident previously evaluated.

Changing Surveillance Requirement 4.1.3.1.4.a from requiring

demonstration of the SDV vent and drain valve operability during a

scram at less than or equal to 50% rod density to a requirement to

perform the testing at shutdown conditions does not significantly

increase the probability or consequences of an accident previously

evaluated. The purpose of the 50% rod density requirement is to

provide a test environment having typical reactor coolant pressure

and temperature conditions. However, the closure time of the vent

and/or drain valves is not affected by pressure and any variations

due to temperature are relatively insignificant. Therefore, testing

from shutdown conditions ensures the safety functions of the vent

and drain valves are met. Also, the proposed change does not affect

system design or operation. Therefore, the proposed change does not

significantly increase the probability or consequences of an

accident previously evaluated.

The deletion of Surveillance Requirement 4.1.3.1.4.b, requiring

proper float switch response by verifying float switch actuation

after a scram, does not significantly increase the probability or

consequences of an accident previously evaluated. Design changes

eliminated the high differential pressure experienced by the float

switches after a scram and provide redundant level measuring

instrumentation. Differential pressure gauges were added in addition

to the float switches to provide the RPS [reactor system protection]

logic with a diverse and redundant means of measuring SDIV [SDV]

level. The changes have resulted in no crushed ball floats at SSES

after a scram. The proposed change will have a negligible effect

[on] the reliability of the system and therefore, does not

significantly increase the probability or consequences of an

accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed change to Technical Specification 3.1.3.1.d, which

extends the restoration time for one inoperable vent and/or one

inoperable drain valve from 24 hours to 7 days, does not create the

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

previously evaluated. Extending the restoration time to 7 days does

not change the design purpose or operation of the SDV valves.

Therefore, the change is bounded by the existing accident analysis.

Changing Surveillance Requirement 4.1.3.1.4.a from requiring

demonstration of the SDV vent and drain valve operability during a

scram at less than or equal to 50% rod density to a requirement to

perform the testing at shutdown conditions does not create the

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

previously evaluated. Performing the surveillance from shutdown

conditions, as recommended in NUREG 1366, ensures that the

operability of the SDV is maintained. No new failure modes are

introduced by the change and the change is bounded by the existing

accident analysis.

The deletion of Surveillance Requirement 4.1.3.1.4.b, requiring

proper float switch response by verifying float switch actuation

after a scram, does not create the possibility of a new or different

kind of accident from any accident previously evaluated. The role of

the ball floats to sense increases in SDV water level is not

affected by the change. In the unlikely event that crushing of the

ball float were to occur, redundant level measuring devices would

maintain system function and the existing 92 day surveillance

activity would identify a ball float failure.

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

margin of safety.

The proposed change to Technical Specification 3.1.3.1.d, which

extends the restoration time for one inoperable vent and/or one

inoperable drain valve from 24 hours to 7 days, does not involve a

significant reduction in a margin of safety. A 7 day completion

time, as identified in NUREG 1433, is reasonable given the level of

redundancy in the lines and the low probability of a scram occurring

while the valve is inoperable and the line is not isolated. Also, a

separate Technical Specification (3.1.3.1.e) addresses the potential

for two inoperable vent or drain valves in the same line.

Changing Surveillance Requirement 4.1.3.1.4.a from requiring

demonstration of the SDV vent and drain valve operability during a

scram at less than or equal to 50% rod density to a requirement to

perform the testing at shutdown conditions does not involve a

significant reduction in a margin of safety. The change maintains

the intent of S.R. 4.1.3.1.4.a by performing equivalent testing at

the same frequency. The change increases the margin of safety by

eliminating potential future scrams taken to meet S.R. 4.1.3.1.4.a.,

thus reducing the potential for safety challenges.

The deletion of Surveillance Requirement 4.1.3.1.4.b, requiring

proper float switch response by verifying float switch actuation

after a scram does not involve a significant reduction in a margin

of safety. Design changes have enhanced the SDV design and provide a

redundant and diverse means of monitoring SDV level. Also, a 92-day

surveillance (S.R. 4.3.1.1) ensures the operability of the float

switch.

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: Mohan Thadani (Acting).

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: December 30, 1993, June 3, 1994, and

August 25, 1994.

Description of amendment requests: The licensee proposes to replace

the current Technical Specifications (TS) with a set of TS based on the

CE Owners Group Improved Standard Technical Specifications issued by

the NRC staff as NUREG-1432 in September 1992. The adoption of Owners

Group-approved TS is part of an industry-wide initiative to standardize

and improve TS. San Onofre, Units 2 and 3, is the lead plant for

adoption of the CE Owners Group standardized TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relocates requirements and surveillances for

structures, systems, components or variables which did not meet the

criteria for inclusion in Technical Specifications. The affected

structures, systems, components or variables are not assumed to be

initiators of analyzed events and are not assumed to mitigate

accident or transient events. The requirements and surveillances for

these affected structures, systems, components or variables will be

relocated from the Technical Specifications to the Licensee

Controlled Specifications or the UFSAR [updated final safety

analysis report]. These operability requirements and surveillances

will continue to be maintained pursuant to 10 CFR 50.59. In

addition, the affected structures, systems, components or variables

are addressed in existing surveillance procedures which are

controlled by 10 CFR 50.59. The reformatting and rewording process

involves no technical changes to plant design or operations.

Therefore, this change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration

of the plant (no new or different type of equipment will be

installed) or changes in parameters governing normal plant

operation. The proposed change will not impose any different

requirements and adequate control of information will be maintained.

Thus, this change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

of safety?

The proposed change will not reduce a margin of safety because

it has no impact on any safety analyses assumptions. In addition,

the affected structure, system, component or variable requirements

and surveillances are the same as the existing Technical

Specifications. Since any future changes to these requirements in

the Licensee Controlled Specifications or UFSAR in the surveillance

procedures will be evaluated per the requirements of 10 CFR 50.59,

no significant reduction in a margin of safety will be allowed.

Therefore, this 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, California 92713.

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

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

NRC Project Director: Theodore R. Quay.

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

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

Alabama

Date of amendments request: August 17, 1994.

Description of amendments request: The proposed amendment would

modify Surveillance Requirement 4.3.1.2, Reactor Trip System

Instrumentation, and Surveillance Requirement 4.3.2.1, Engineered

Safety Feature Actuation System Instrumentation, to eliminate the

periodic pressure sensor response time testing requirements. The

Surveillance Requirements would indicate that the total channel

response time will be periodically ``verified'' instead of ``tested''.

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 license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This change to the Technical Specifications does not result in a

condition where the design, material, and construction standards

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

and ESFAS instrumentation is being used; the time response

allocations/modeling assumptions in the FSAR Chapter 15 analyses are

still the same; only the method of verifying time response is

changed. Periodic calibration of these pressure and differential

pressure instruments will detect significant degradation in the

sensor response characteristic and assure equipment operability.

Flushing of selected sensing lines during each refueling outage as

recommended by NUREG/CR5851, ``Long Term Performance and Aging

Characteristics of Nuclear Plant Pressure Transmitters,'' will

mitigate sensing line response time degradation due to blockage that

noise analysis testing techniques would have previously detected.

The proposed change will not modify any system interface and could

not increase the likelihood of an accident since these events are

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

degrade or prevent actions or alter any assumptions previously made

in evaluating the radiological consequences of an accident described

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

any increase in the probability or consequences of an accident

previously evaluated.

2. The proposed license amendment does not create the

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

previously evaluated.

This change does not alter the performance of the pressure and

differential pressure transmitters and switches used in the plant

protection systems. All sensors will still have response time

verified by test before placing the sensors in operational service

and after any maintenance that could affect response time. Changing

the method of periodically verifying instrument response for certain

sensors from time response testing to calibration will not create

any new accident initiators or scenarios. Implementation of the

proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in margin of safety.

This change does not affect the total system response time

assumed in the safety analysis. The periodic system response time

verification method for selected pressure and differential pressure

sensors is modified to allow use of actual test data or engineering

data. The method of verification combined with sensing line

preventative maintenance provides assurance that the total system

response is within the time limit defined in the safety analysis,

since calibration tests will detect any degradation which might

significantly affect pressure sensor response time and periodic

sensing line flushing will minimize the potential for long-term

buildup of contaminants which may impact sensing line response.

Based on the above, it is concluded that the proposed license

amendment request does not result in a reduction in margin with

respect to plant safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: James H. Miller, III, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201.

NRC Project Director: David B. Matthews.

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

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

Date of amendment request: March 30, 1994 (TS 318).

Description of amendment request: The proposed amendment consists

of six parts, which the licensee has designated ``A'' through ``F'' in

the analysis of no significant hazards considerations presented below.

The six parts are:

A. Mechanical pressure and differential pressure indicating

switches in the Reactor Protection System (RPS) and Emergency Core

Cooling System (ECCS) are being replaced with an Analog Transmitter/

Trip System (ATTS) for the Browns Ferry Nuclear Plant (BFN) Unit 3.

B. The BFN Units 1 and 3 reactor vessel water level safety limit is

being revised to reflect the analytical limit provided by General

Electric and the Level 1 Low Reactor Vessel Water Level setpoint is

being revised to provide a more conservative limit.

C. For BFN Unit 2, RPS and ECCS instrument identifiers are being

added or corrected to enhance useability of the Technical

Specifications. These changes do not reflect a change in equipment,

operation of the associated system, or the safety function of that

system.

D. For BFN Unit 2, Reactor High Water Level, Reactor Core Isolation

Cooling (RCIC) and High Pressure Coolant Injection (HPCI) Turbine Steam

Line High Flow, and Drywell Pressure instrumentation calibration

frequencies and functional test descriptions are being revised to

reflect current calculations and test methods. These changes do not

reflect a change in equipment, operation of the associated system, or

the safety function of that system.

E. For BFN Units 1, 2, and 3, the differential pressure

instrumentation, which actuates the pressure suppression chamber-

reactor building vacuum breakers, calibration frequency is being

revised. In addition, tables that specify the minimum number of

instrument channels per trip system, function, trip level setting,

actions required, remarks, functional test, and instrument check are

being added.

F. Corrects the capitalization of terms used on the affected BFN

Units 1, 2, and 3 TS pages in order to conform with the current TS

Definitions section. This part also corrects spelling and

capitalization of other words on the same pages.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Part A: The Unit 3 modification, which involves the installation

of an Analog Transmitter/Trip System (ATTS), replaces older devices

with devices of more modern design that perform the same function.

The initiation of control rod insertion to mitigate a design

basis accident is contained in Chapter 14 of the BFN Final Safety

Analysis Report (FSAR). There is no change in design bases,

protective function (initiation of control rod insertion),

redundancy, setpoints, or logic associated with the installation of

the ATTS. The consequences of a failure of this equipment are no

different than that of the original equipment. Since there is no

change in any protective functions, nor the creation of any new

operational conditions, the proposed amendment does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

Part B: The revision to the Units 1 and 3 reactor vessel water

level safety limit and the Level 1 low reactor vessel water level

setpoint do not reflect any change in plant equipment. The safety

limit is being changed to reflect the actual analytical safety limit

calculated by General Electric.

The Level 1 low reactor vessel water level trip initiates the

Core Spray and Low Pressure Coolant Injection Systems and isolates

the Main Steam lines. These actions are taken to mitigate the

consequences of a Loss of Coolant Accident. The change in the

setpoint affects the timing of the operation of equipment necessary

to mitigate the consequences of an accident. A setpoint calculation

has been generated which ensures these safety functions are

initiated in accordance with the design basis accident analysis

presented in Chapter 14 of the Browns Ferry FSAR. Therefore, the

proposed amendment does not involve a significant increase in the

probability or consequences of any accident previously evaluated.

Part C: The addition or correction of Unit 2 instrument

identifiers is administrative in nature and does not reflect any

modification to plant equipment. These administrative changes do not

reflect any change to any precursor for the design basis events or

operational transients analyzed in the Browns Ferry FSAR. There is

also no change to any protective function or mitigating action for

the design basis events or operational transients analyzed in the

Browns Ferry FSAR. Therefore, the probability or consequences of an

accident previously evaluated is not significantly increased.

Part D: The change in Unit 2 reactor high water level and

Reactor Core Isolation Cooling (RCIC) instrumentation functional

test descriptions reflects the equipment currently installed and the

functional tests currently being performed.

The changes in calibration frequencies are being made to reflect

current setpoint calculations. There are no modifications to plant

equipment or changes in instrument setpoints associated with these

changes. The calibration frequencies specified by the current

setpoint calculations ensure that the associated safety functions

are initiated in accordance with the design basis accident analysis

presented in Chapter 14 of the Browns Ferry Final Safety Analysis

Report (FSAR). Therefore, the probability or consequences of an

accident previously evaluated is not significantly increased.

Part E: The changes in Units 1, 2, and 3 calibration frequency

for the differential pressure instrumentation, which actuates the

pressure suppression chamber-reactor building vacuum breakers, is

being made to reflect current setpoint calculations. The specified

minimum number of instrument channels per trip system, function,

trip level setting, actions required, remarks, functional test, and

instrument check reflect current operational requirements. There are

no modifications to plant equipment or changes in instrument

setpoints associated with these changes. The calibration frequencies

specified by the current setpoint calculations ensure that the

associated safety functions are initiated in accordance with the

design basis accident analysis presented in Chapter 14 of the Browns

Ferry Final Safety Analysis Report (FSAR). Therefore, the

probability or consequences of an accident previously evaluated is

not significantly increased.

Part F: The proposed correction of the capitalization of terms

in order to conform with the current TS Definitions section is

administrative in nature and does not reflect any modification to

plant equipment. The correction of spelling and capitalization of

other words on the same pages is also administrative in nature and

does not reflect any modification to plant equipment. These

administrative changes do not reflect any change to any precursor

for the design basis events or operational transients analyzed in

the Browns Ferry FSAR. There is also no change to any protective

function or mitigating action for the design basis events or

operational transients analyzed in the Browns Ferry FSAR. Therefore,

the probability or consequences of an accident previously evaluated

is not significantly increased.

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

new or different kind of accident from any accident previously

evaluated.

Part A: The installation of the ATTS replaces older devices with

devices of more modern design that perform the same function. No new

control functions are added. No credible equipment failure modes or

single failure are introduced which could result in the inability of

redundant safety components or systems to perform their safety

functions in accordance with the design basis accident analysis

presented in Chapter 14 of the Browns Ferry FSAR. Therefore, the

proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Part B: The revision to the Units 1 and 3 reactor vessel water

level safety limit and the Level 1 low reactor vessel water level

setpoint do not reflect any change in plant equipment. The safety

limit is being changed to reflect the actual analytical safety limit

calculated by General Electric.

The change in the Level 1 low reactor vessel water level

setpoint affects the timing of the operation of equipment necessary

to mitigate the consequences of an accident. No new failure modes or

system interactions are introduced. The same protection functions

will still occur at the Level 1 low reactor water level setpoint.

Therefore, the proposed amendment does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

Part C: The addition or correction of Unit 2 instrument

identifiers is administrative in nature and does not reflect any

modification to plant equipment. The correction of instrument

identifiers does not require new system alignments, modifications,

or changes in operating procedures. Therefore, no new external

threats, system interactions, release pathways, equipment failure

modes, or types of operator errors are created. Therefore, the

proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Part D: The change in Unit 2 reactor high water level and RCIC

instrumentation functional test descriptions reflects the equipment

currently installed and the functional tests currently being

performed. The changes in calibration frequencies are being made to

reflect current setpoint calculations. There are no modifications to

plant equipment or changes in instrument setpoints associated with

these changes. No new failure modes or system interactions are

introduced. The same protection functions will still occur at the

same setpoints. Therefore, the proposed amendment does not create

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

accident previously evaluated.

Part E: The changes in Units 1, 2, and 3 calibration frequency

for the differential pressure instrumentation, which actuates the

pressure suppression chamber-reactor building vacuum breakers, is

being made to reflect current setpoint calculations. The specified

minimum number of instrument channels per trip system, function,

trip level setting, actions required, remarks, functional test, and

instrument check reflect current operational requirements. There are

no modifications to plant equipment or changes in instrument

setpoints associated with these changes. No new failure modes or

system interactions are introduced. The same protection functions

will still occur at the same setpoints. Therefore, the proposed

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

of accident from any accident previously evaluated.

Part F: The proposed correction of the capitalization of terms

in order to conform with the current TS Definitions section is

administrative in nature and does not reflect any modification to

plant equipment. The correction of spelling and capitalization of

other words on the same pages is also administrative in nature and

does not reflect any modification to plant equipment. The correction

of spelling and capitalization does not require new system

alignments, modifications, or changes in operating procedures.

Therefore, no new external threats, system interactions, release

pathways, equipment failure modes, or types of operator errors are

created. Therefore, the proposed amendment does not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

Part A: The installation of the ATTS replaces older devices with

devices of more modern design that perform the same function. The

replacement equipment will improve reliability, accuracy and

response times. There are no changes in the systems' design basis,

protective function, or logic arrangement. Instrument setpoints and

calibration frequencies are supported by Unit 3 specific

calculations. Therefore, the proposed amendment does not involve a

significant reduction in the margin of safety.

Part B: The revision to the Units 1 and 3 reactor vessel water

level safety limit and the Level 1 low reactor vessel water level

setpoint do not reflect any change in plant equipment. The safety

limit is being changed to reflect the actual analytical safety limit

calculated by General Electric.

The change in the Level 1 low reactor vessel water level

setpoint is supported by a Unit 3 specific setpoint calculation that

has been performed in accordance with the methodology endorsed by

Regulatory Guide 1.105, Instrument Setpoints for Safety Related

Systems. Therefore, the proposed amendment does not involve a

significant reduction in the margin of safety.

Part C: The addition or correction of Unit 2 instrument

identifiers is administrative in nature and does not reflect any

modification to plant equipment. Therefore, the proposed amendment

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

Part D: The change in Unit 2 reactor high water level and RCIC

instrumentation functional test descriptions reflects the equipment

currently installed and the functional tests currently being

performed. The changes in calibration frequencies are being made to

reflect Unit 2 specific setpoint calculations. These calculations

have been performed in accordance with the methodology endorsed by

Regulatory Guide 1.105. There are no modifications to plant

equipment or changes in instrument setpoints associated with these

changes. Therefore, the proposed amendment does not involve a

significant reduction in the margin of safety.

Part E: The changes in Units 1, 2, and 3 calibration frequency

for the differential pressure instrumentation, which actuates the

pressure suppression chamber-reactor building vacuum breakers, is

being made to reflect current setpoint calculations. The specified

minimum number of instrument channels per trip system, function,

trip level setting, actions required, remarks, functional test, and

instrument check reflect current operational requirements. The

setpoint calculations have been performed in accordance with the

methodology endorsed by Regulatory Guide 1.105. There are no

modifications to plant equipment or changes in instrument setpoints

associated with these changes. Therefore, the proposed amendment

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

Part F: The proposed correction of the capitalization of terms

in order to conform with the current TS Definitions section is

administrative in nature and does not reflect any modification to

plant equipment. The correction of spelling and capitalization of

other words on the same pages is also administrative in nature and

does not reflect any modification to plant equipment. Therefore, the

proposed amendment 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: Athens Public Library, South

Street, Athens, Alabama 35611.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

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

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

Date of amendment request: March 31, 1994 (TS 339).

Description of amendment request: The proposed amendment consists

of five parts, which the licensee has designated ``A'' through ``E'' in

the discussion of no significant hazards considerations, below. These

five parts are:

A. For the Browns Ferry Nuclear Plant (BFN) Units 1 and 3, improve

operating flexibility by expanding the allowable operating domain. This

expansion is accomplished by revising the equations for the flow-biased

Average Power Range Monitor (APRM) flux scram trip setting and the APRM

rod block trip setting.

B. Revises the Rod Block Monitor (RBM) limiting conditions for

operation (LCOs) to require two RBM channels to be operable if the

plant is operating with low thermal margins.

C. Miscellaneous editorial changes to the BFN Units 1, 2, and 3

technical specifications (TS).

D. Revises the BFN Units 1, 2, and 3 TS to delete the specific

value for the rated loop recirculation flow rate.

E. Revises the BFN Units 1, 2, and 3 TS to relocate the specific

equations for the APRM rod block and RBM upscale setpoint equations

from the TS to 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. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Part A: The proposed change will permit expansion of the current

allowable power/flow operating region to allow operation in the

extended load line limit (ELLL) region. Operation of BFN Units 1 and

3 in the ELLL region will not increase the probability of any

accident previously evaluated since the Average Power Range Monitor

(APRM) system and flow-biased scram setpoint are not identified as

initiators of any design basis accidents or transients.

Additionally, no credit is taken for the APRM flow biased scram in

any accident or transient analyses. Therefore, the proposed change

can not significantly increase the probability of an accident

previously evaluated.

TVA's analysis of operation in the ELLL region verified that the

consequences of previously evaluated accidents are within the

acceptance criteria of the licensing basis. Therefore, the proposed

change does not involve an increase in the consequences of an

accident previously evaluated.

Part B: The proposed change does not increase challenges or

create any new challenges to safety-related systems or equipment, or

other equipment whose failure could cause an accident. The proposed

change does not change the function of the rod block monitor (RBM)

subsystem. The RBM subsystem will continue to block control rod

withdrawal to ensure that fuel safety limits are protected. The

revised RBM limiting conditions for operation and surveillance

requirements provide increased assurance that the RBM will function

to ensure that fuel safety limits are protected. Therefore, the

proposed change does not involve an increase in the probability of

an accident previously evaluated.

The revised RBM operability and surveillance requirements

provide increased assurance that the RBM will block control rod

withdrawal to ensure that fuel safety limits are protected.

Accordingly, operation of BFN Units 1 and 3 with the revised RBM

upscale setpoint does not involve an increase in the consequences of

an accident previously evaluated.

Part C: The miscellaneous editorial changes do not affect any

plant operations, equipment, or any safety-related activity. These

changes increase the probability that the specifications will be

correctly interpreted by adding clarifying information and/or

correcting errors. Therefore, these editorial changes do not involve

an increase in the probability or consequences of an accident

previously evaluated.

Part D: The proposed change will delete the specific value for

the rated loop recirculation flow rate found in the limiting safety

system settings. This flow rate is in the TS to provide additional

information, and is not a TS requirement. The proposed change does

not change the limiting safety system settings or alter the method

for calculating the settings. The proposed change does not affect or

change operation of the plant, plant equipment, or any safety-

related equipment. The proposed change does not change the APRM rod

block or trip settings, the method or frequency of calibration of

the APRM flow biased network, or any other operational features of

the APRM system. The proposed change will only delete an incorrect

flow rate from the TS. Therefore, the proposed change does not

involve an increase in the probability or consequences of an

accident previously evaluated.

Part E: The proposed change will remove specific equations for

the APRM rod block and RBM upscale trip setpoints from the TS and

relocate them to the Core Operating Limits Report (COLR). Removing

these equations from the TS does not affect or change the APRM and

RBM subsystems or the functions of these systems. The proposed

change does not affect or change operation of the plant, plant

equipment, or any safety-related equipment. Accordingly, the

proposed change does not involve an increase in the probability of

an accident previously evaluated.

Removing the specific rod block equations from the TS does not

change the requirements to comply with the limits of these equations

during plant operations, since the TS will reference the COLR as the

source of the equations. The actions to be taken in the event of

noncompliance with the COLR-specified equations will also remain

unchanged. Both the APRM rod block and RBM subsystems will continue

to block control rod withdrawal to prevent reactor power from

increasing to excess levels and to ensure that applicable limits of

the plant safety analysis are met. Additionally, in accordance with

the requirements of TS 6.9.1.7, these equations will continue to be

developed using NRC-approved methodologies and will continue to

ensure that applicable safety limits are protected. Therefore, the

proposed change does not involve an increase in the consequences of

an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Part A: Operation of BFN Unit 1 and 3 in the ELLL region does

not create any new failure mode or sequence of events that can lead

to an accident of a different type than any previously evaluated.

Operation in the ELLL region does not increase challenges or create

any new challenges to safety-related systems or equipment, or other

equipment whose failure could cause an accident. Changing the

equation for the flow-biased APRM scram trip setpoint does not

change the function of the APRM subsystem. The APRM scram trip

setpoint will continue to initiate a scram to ensure that the fuel

safety limit is not exceeded. Therefore, the proposed change does

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

from any accident previously evaluated.

Part B: The proposed change to the RBM operability and

surveillance requirements does not create any new failure mode or

sequence of events that can lead to an accident of a different type

than any previously evaluated. The proposed change does not increase

challenges or create any new challenges to safety-related systems or

equipment, or other equipment whose failure could cause an accident.

The proposed change does not change the function of the RBM

subsystem. The RBM subsystem will continue to block control rod

withdrawal to ensure that fuel safety limits are protected.

Therefore, the proposed change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Part C: The miscellaneous editorial changes do not affect any

plant operations, equipment, or any safety-related activity. These

changes increase the probability that the specifications will be

correctly interpreted by adding clarifying information and/or

correcting errors. Therefore, these editorial changes do not create

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

accident previously evaluated.

Part D: The proposed change will delete the specific value for

the rated loop recirculation flow rate found in a factor used to

calculate limiting safety system settings for the APRM rod block and

trip settings. The proposed change does not change the limiting

safety system settings or alter the method for calculating the

settings. The proposed change does not affect or change operation of

the plant. The proposed change does not change the APRM rod block or

trip settings, the method or frequency of calibration of the APRM

flow biased network, or any other operational features of the APRM

system. The proposed change will only delete an incorrect flow rate

that is required to be calculated and maintained outside of the TS.

Since there will be no change in plant operations, the proposed

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

accident from any accident previously evaluated.

Part E: Removal of the APRM rod block and RBM upscale setpoint

equations does not change or affect any safety-related plant

equipment or its functions; or any equipment, systems, or setpoints

designed to prevent or mitigate accidents. Removing these rod block

equations does not create any new challenges to safety-related

systems or equipment, or other equipment whose failure could cause

an accident; and does not change the function and manner of

operation of the APRM or RBM subsystems. The APRM and RBM subsystems

will continue to block control rod withdrawal to prevent reactor

power from increasing to excess levels and to ensure that fuel

safety limits are protected. Therefore, the proposed change does not

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

any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

Part A: Operation of BFN Units 1 and 3 in the ELLL region does

not affect the ability of the plant safety-related trips or

equipment to perform their intended functions. Operation in the ELLL

region will not cause any significant increase in offsite radiation

doses resulting from any analyzed event. Although this change

increases the APRM flow-biased scram equation, no credit is taken

for this equation in the accident analyses. These analyses assume

that transient events initiated from less than rated conditions are

terminated by the fixed 120% flux scram or other safety-grade scram

signals. These signals are not affected by the proposed change.

Additionally, as noted above, TVA's analysis of operation in the

ELLL region determined that the consequences of previously evaluated

accidents remain within the acceptance criteria of the licensing

basis. Therefore, this change does not involve a reduction in a

margin of safety.

Part B: The proposed change does not change the function of the

RBM system. The RBM system will continue to block control rod

withdrawal to ensure that fuel safety limits are protected. The

proposed change does not affect plant operation, design, or any

safety-related activity or equipment. The proposed change does not

affect or change any margin of safety. The proposed change will

actually increase the margin of safety by providing more

conservative operability and surveillance requirements for the RBM

subsystem. Therefore, the proposed change does not involve a

reduction in a margin of safety.

Part C: The miscellaneous editorial changes do not affect plant

operation, design, or any safety-related activity or equipment.

These changes increase the probability that the specifications will

be correctly interpreted by adding clarifying information and/or

correcting errors. Therefore, these changes do not involve a

reduction in a margin of safety.

Part D: The proposed change will delete the specific value for

the rated loop recirculation flow rate found in the limiting safety

system settings. The proposed change does not change the limiting

safety system settings or alter the method for calculating the

settings. The proposed change does not affect or change any margin

of safety. The proposed change does not alter the APRM rod block or

trip settings, nor does it change the combinations of power and flow

conditions which could produce the APRM flow biased rod block and

scram trips. Furthermore, the value for rated loop recirculation

flow rate will continue to be contained in plant procedures which

are controlled by the 10 CFR 50.59 process. Therefore, the proposed

change does not involve a reduction in a margin of safety.

Part E: The proposed change to remove the APRM rod block and RBM

upscale setpoint equations does not change the equations or alter

the method for calculating the equations. The proposed change does

not change or affect any safety-related plant equipment or its

functions; or any equipment, systems, or setpoints designed to

prevent or mitigate accidents. Removing these rod block equations

does not create any new challenges to safety-related systems or

equipment, or other equipment whose failure could cause an accident;

and does not change the function and manner of operation of the APRM

or RBM subsystems. The requirements of TS 6.9.1.7 will continue to

ensure that these equations are developed using NRC-approved

methodology, and are consistent with applicable limits of the plant

safety analysis. 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: Athens Public Library, South

Street, Athens, Alabama 35611.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: August 9, 1994.

Brief description of amendments: The proposed amendments would

revise the technical specifications by eliminating the high negative

neutron flux rate trip function.

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 deletion of the High Negative Neutron Flux Rate trip does

not adversely affect the probability of any accident. In fact, the

deletion of this reactor trip is expected to reduce the probability

of inadvertent reactor trips during surveillance testing.

The only previously evaluated accidents whose consequences could

be potentially affected by this change are the dropped rod events.

Presently the High Negative Neutron Flux Rate trip function responds

to these events by initiating a reactor trip. Analyses of these

events, using currently licensed analysis methodologies have

demonstrated that this trip function is not necessary. Although the

scenarios for the various rod drop events without this trip function

differ from the existing event scenarios, the unit will either be

safely shutdown or return to an acceptable reactor power level, and

as before, DNB does not occur. Therefore, the proposed changes would

not have a significant effect on the consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes remove a feature from the Reactor Trip

System that is a potential source of inadvertent or unnecessary

reactor trips. While the changes delete an accident mitigation

feature, they do not add new hardware to the units and do not change

plant operations; consequently, no new failure modes are introduced.

Therefore, the removal of this trip function cannot create the

possibility of an [sic] new or different kind of accident from any

accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

Because the changes only delete an unnecessary reactor trip and

do not actually alter the plant, the proposed changes do not affect

the failure values for any system or component. Accident analyses

have shown that all relevant, existing event acceptance criteria

have been satisfied without taking credit for the deleted reactor

trip; therefore, the event acceptance criteria are not being

revised. Because neither the failure values nor the acceptance

criteria are affected, the proposed changes have no affect on the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.

NRC Project Director: William D. Beckner.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: August 12, 1994.

Brief description of amendments: The proposed amendments would

revise the Administrative Controls section of the technical

specifications to reflect changes to the licensee's organization,

change the submittal due date for the Monthly Operating Report, and

delete provisions that will be relocated to other controlled documents.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes are administrative in nature and reflect

new organizational position titles, reassigned managerial

responsibilities, a change in the required submittal due date of the

Monthly Operating Report, and the relocation of selected Technical

Specification Administrative Controls to licensee controlled

documents. The proposed organizational changes include manager

position title updates as follows: ``Group Vice President, Nuclear

Engineering and Operations'' changed to ``Group Vice President,

Nuclear Production''; ``Vice President, Nuclear Operation'' changed

to ``Vice President of Nuclear Operations''; and ``Shift

Supervisor'' changed to ``Shift Manager''. Manager responsibilities

remain unchanged except the annual management directive concerning

the Shift Supervisor's control room command function which is

reassigned to the Group Vice President, Nuclear Production.

Additionally, responsibilities of the Vice President, Nuclear

Operations and Plant Manager are proposed for reassignment to either

the Vice President of Nuclear Operations or the Plant Manager. In

all cases, responsibilities will continue to be assigned to

appropriately qualified individuals.

The proposed change in the submittal due date of the Monthly

Operating Report is to provide additional time to facilitate data

compilation and report preparation, review and approval for two

operating units. This change does not alter any data or information

already reported.

The proposed administrative control relocations include details

of minimum shift operations crew staffing, method of onsite and

offsite review and audit, selected reportable event actions, review

of security and emergency plans and associated implementing

procedures, requirements for record retention, and program elements

for radiation protection, process control and radiological

environmental monitoring. These administrative controls are

addressed by other regulatory requirements and are relocated to

other licensee documents (i.e., Final Safety Analysis Report,

Emergency Plan, Security Plan, Offsite Dose Calculation Manual)

which have adequate change control to ensure that intended plant

design/safety functions will be maintained. No design basis

accidents are affected by these proposed administrative changes as

they do not impact nor affect accident analysis assumptions.

Therefore, accident analyses assumptions are preserved and there is

no change in the probability or consequences of any previously

evaluated accident.

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 update organizational position titles

and reassign responsibilities will not delete any responsibility/

function already designated in the Technical Specifications. All

given management activities will continue to be performed by

qualified individuals. The proposed change in the submittal due date

of the Monthly Operating Report does not alter any data or

information already reported. The administrative control relocations

retain adequate regulatory basis to ensure that intended plant

design/safety functions will be maintained. These changes are

administrative in nature and do not affect the design or operation

of any system, structure, or component in the plant. Accordingly, no

new failure modes have been defined for any plant system or

component important to safety, nor have any new initiating events

been identified as a result of the proposed changes. Therefore, the

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

kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes involve the Administrative Controls Section

of the CPSES Units 1 and 2 Technical Specifications and provide for

updating organizational position titles, reassigning managerial

responsibilities/functions, changing the required submittal due date

for the Monthly Operating Report, and relocating administrative

controls to other controlled licensee documents. These changes are

administrative in nature and do not directly affect any protective

boundaries nor impact the safety limits for the protective

boundaries. The relocated requirements retain adequate regulatory

basis for continued proper administrative review and plant

configuration control to ensure that actions prescribed in plant

operating procedures are maintained so as not to impact the plant's

margin of safety. 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 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 Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.

NRC Project Director: William D. Beckner.

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

Callaway County, Missouri

Date of amendment request: August 4, 1994.

Description of amendment request: This proposed amendment would

reflect changes in the boron dilution accident analysis to address

Boron Dilution Mitigation System (BDMS) time delays, the BDMS actuation

setpoint uncertainty, and concerns regarding the applicability of the

assumed inverse count rate ratio (ICRR) curve. The Technical

Specifications include:

a. Changing Bases page 2-8, note** of Table 3.3-1, and note 12 of

Table 4.3-1 to reflect ``flux multiplication'' rather than ``flux

doubling;''

b. Revising note 9 of Table 4.3-1 to reflect the revised setpoint;

c. Revising Bases page 3/4 4-1 to reflect new analysis assumptions.

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.

The actuation setpoint decrease and administrative controls to

isolate dilution sources if no reactor coolant loop is in operation

during Modes 3-5 reflect the new analysis assumptions. The

initiating events are presented in FSAR Section 15.4.6. The proposed

changes affect only the time required for BDMS to mitigate the event

and do not affect the probability of any event initiators.

Overall protection system performance will remain within the

bounds of the accident analyses documented in FSAR Chapter 15, WCAP-

10961-P, and WCAP-11883 since no hardware changes are proposed.

The BDMS will continue to function in a manner consistent with

the above analysis assumptions and the plant design basis. As such,

there will be no degradation in the performance of nor an increase

in the number of challenges to equipment assumed to function during

an accident situation.

These Technical Specification revisions do not involve any

hardware changes nor do they affect the probability of any event

initiators. There will be no change to normal plant operating

parameters or accident mitigation capabilities. Therefore, there

will be no increase in the probability of any accident occurring due

to the revised analysis.

The results of this new analysis indicate that there is

sufficient time for BDMS action to prevent a loss of plant shutdown

margin. Since plant shutdown margin is not lost, the minimum DNBR

remains well above the safety analysis limit values, no

overpressurization occurs and, therefore, there are no fuel

failures. The Technical Specification limits on shutdown margin in

Modes 3-5 will be met. The conclusions of NRC Generic Letter 85-05

and NSAC-183 remain valid (i.e., that gradual boron dilution events

are self-limiting due to inherent reactivity feedback mechanisms).

Given the above, there will be no increase in the consequences of

any accident.

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

accident from any previously evaluated.

As discussed above, there are no hardware changes associated

with these Technical Specification revisions nor are there any

changes in the method by which any safety-related plant system

performs its safety function. The normal manner of plant operation

is unaffected.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of these changes. There will be no adverse effect or challenges

imposed on any safety-related system as a result of these changes.

Therefore, the possibility of a new or different type of accident is

not created.

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

The results of the new analysis show that there is sufficient

time for BDMS action to prevent a loss of plant shutdown margin.

Since plant shutdown margin is not lost, the minimum DNBR remains

well above the safety analysis limit values. The Technical

Specification limits on shutdown margin in Modes 3-5 will be met.

There will be no effect on the manner in which safety limits or

limiting safety system settings are determined nor will there be any

effect on those plant systems necessary to assure the accomplishment

of protection functions. There will be no impact on DNBR limits,

FQ, F-delta-H, LOCA PCT, peak local power density, or any other

margin of safety.

Based upon the preceding information, it has been determined

that the proposed changes to the Technical Specifications do not

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 accident previously evaluated,

or involve a significant reduction in a margin of safety. Therefore,

it is concluded that the proposed changes meet the requirements of

10CFR50.92(c) and does 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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: John N. Hannon.

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

Callaway County, Missouri

Date of amendment request: August 4, 1994.

Description of amendment request: The proposed amendment would

replace Technical Specification (TS) 3/4.6.2.2, Spray Additive System,

with a new TS 3/4.6.2.2 entitled Recirculation Fluid pH control (RFPC)

System. The associated TS Surveillance Requirements and the Bases would

be revised. In addition, the Bases for the Refueling Water Storage Tank

(RWST) System will be revised.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Overall protection system performance will remain within the

bounds of the accident analyses documented in FSAR Chapter 15, WCAP-

10961-P, and WCAP-11883.

The accidents evaluated in the FSAR that could be affected by

this proposed change are those involving the pressurization of the

containment and associated flooding of the containment and

recirculation of this fluid within the ECCS or the Containment Spray

System (i.e., large break LOCA, main steam line break inside

containment, and feedwater line break inside containment). The TSP-C

will dissolve in the containment sump fluid resulting from these

accidents raising the pH of the fluid, which would initially be

greater than or equal to 4.0 but less than 7.0 during the injection

phase of containment spray operation. The equilibrium spray pH

during the recirculation phase resulting from this change will be

greater than or equal to 7.1. The pH range for the spray will be

bounded by the range of 4.0 to 11.0 in the current FSAR Section

3.11(B) for the postulated spray solution environment. Since the

resulting pH level will be closer to neutral using the TSP-C instead

of NaOH, post-LOCA corrosion of containment components will not be

increased. The results of the current, post-LOCA hydrogen generation

calculation will remain bounding. There will not be an adverse

radiation dose effect on any safety-related equipment. Thus, the

potential for failures of the ECCS or safety-related equipment

following a LOCA will not be increased as a result of the proposed

change. The radiological consequences of changing from NaOH to TSP-C

were reanalyzed using the current NRC methodology presented in

Revision 2 of the Standard Review Plan (NUREG-0800) Section 6.5.2.

This reanalysis indicates that the proposed change would result in

reduced control room doses. Offsite doses would remain less than

those currently reported in FSAR Table 15.6-8. The offsite and

control room doses will continue to meet the requirements of 10 CFR

100, 10 CFR 50 Appendix A GDC 19, SRP 15.6.5.II, and SRP 6.4.II. The

dose reanalysis, combined with knowledge gained from recent studies

on the behavior of iodine in the post-LOCA environment, demonstrates

that the deletion of the Spray Additive System and replacement with

a sump pH control system using TSP-C will not increase the reported

radiological consequences of a postulated LOCA. The proposed new pH

control system will provide satisfactory retention of iodine in the

sump water, as well as provide adequate pH control to minimize the

potential of chloride-induced stress corrosion cracking of

austenitic stainless steel components.

The Containment Spray System will continue to function in a

manner consistent with the plant design basis. There will be no

degradation in the performance of nor an increase in the number of

challenges to equipment assumed to function during an accident

situation.

These Technical Specification revisions do not affect the

probability of any event initiators. There will be no change to

normal plant operating parameters, ESF actuation setpoints, or

accident mitigation capabilities. Therefore, these changes will 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.

The new Recirculation Fluid pH Control System is a passive

system, i.e., no operator or automatic action is required to actuate

the system. There are no active components being added whose failure

could prevent the new system from functioning. The only new

components being added are the TSP-C storage baskets. Seismic

requirements have been included in the design to ensure the

structural integrity of the baskets will be maintained during a

seismic event.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of these changes. There will be no adverse effect or challenges

imposed on any safety-related system as a result of these changes.

The use of dry sodium phosphates is allowed for adjustment of the

post-LOCA sump solution pH as discussed in SRP 6.1.1. Trisodium

phosphate has a dissolution rate of 0.7 lbm/ft2-min in water at

160 deg.F (given in WCAP-12477, based on trisodium phosphate in the

form of a solid block with no agitation of the solution). The

quantity of trisodium phosphate chosen will provide a minimum

equilibrium sump pH of 7.1 following dissolution and mixing. No new

equipment performance burdens are imposed' however, there is the

potential for an unlikely, but possible, event in which an initially

concentrated solution of TSP-C occupies the stagnant volume of an

inoperable sump. This situation would not last for long since, as

the recirculated sump fluid is cooled in the RHR heat exchangers,

sufficient buoyancy-driven circulation within containment will

result to displace the stagnant solution and eventually yield a

uniform, equilibrium solution. Therefore, the possibility of a new

or different type of accident is not created.

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

The radiological analysis performed for this proposed change, as

discussed above, shows that there would be no impact on the doses

reported in FSAR Table 15.6-8.

There will be no change to the DNBR Correlation Limit, the

design DNBR limits, or the safety analysis DNBR limits discussed in

Bases Section 2.1.1.

There will be no effect on the manner in which safety limits or

limiting safety system settings are determined nor will there be any

effect on those plant systems necessary to assure the accomplishment

of protection functions. There will be no impact on DNBR limits,

FQ, F-delta-H, LOCA PCT, peak local power density, or any other

margin of safety.

Based upon the preceding information, it has been determined

that the proposed changes to the Technical Specifications do not

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 accident previously evaluated,

or involve a significant reduction in a margin of safety. Therefore,

it is concluded that the proposed changes meet the requirements of

10 CFR 50.92(c) and 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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: John N. Hannon.

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

Project No. 2, Benton County, Washington

Date of amendment request: January 6, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to (1) modify a table notation

which would allow reactor protection channels to be placed in an

inoperable status for a defined period of time to allow conduct of

surveillance testing without requiring entry into the associated

limiting condition for operation (LCO) action requirements, and (2)

delete the channel check requirements for the ``Reactor Steam Dome--

High'' TS surveillance.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below:

Regarding the proposed modifications to the Instrumentation Section

table notations:

1. Does the proposed change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed revised notations to the TS Instrumentation Section

tables clarify the current intent of the table notations by allowing

the conduct of surveillance testing without declaring the affected

channels inoperable for a limited period of time. In addition, the

proposed changes make the wording of the table notations consistent

with the Improved Technical Specifications (ITS). The proposed changes

do not affect the time that the current notations allow a channel to be

placed in an inoperable status, and thus do not affect the probability

of an accident previously evaluated. In addition, the proposed changes

do not affect the current requirement that redundant instrumentation be

operable to accomplish the required function for the channel placed in

an inoperable status. Thus, the proposed changes do not affect the

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes do not affect the operation or availability of

equipment from the current TS, and therefore do not create the

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

previously evaluated.

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

of safety?

The margin of safety affected by the proposed changes is the time

that instrumentation required for plant protection is unavailable to

perform its required function(s). The proposed changes do not affect

the time that the instrumentation is allowed to be in an inoperable

status during surveillance testing, and thus does not affect the

availability of required instrumentation. In addition, the proposed

changes do not affect the current requirement that redundant

instrumentation remains operable, ensuring the required safety function

remains operable during surveillance testing. Thus, the margin of

safety is not affected by the proposed changes.

Regarding the proposed removal of the channel check requirement for

the ``Reactor Vessel Steam Dome Pressure--High'' instrumentation:

1. Does the proposed change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed removal of this channel check requirement would not

affect the availability of the trip function associated with the

``Reactor Vessel Steam Dome Pressure--High'' instrument, since the

status of the channels is monitored by existing annunciators. This

ensures the instruments remain capable of performing their intended

function.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a physical modification to the

facility or its equipment, nor does it result in a change in the

reliability of the equipment. The proposed change does not, therefore,

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

accident previously evaluated.

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

of safety?

The margin of safety associated with not performing channel checks

on the ``Reactor Vessel Steam Dome Pressure--High'' instrument is

defined by the plant response to accidents that depend on this

instrument, which in turn is dependent on the likelihood of

availability of the trip function associated with this instrument. The

proposed change does not affect the availability of this instrument,

since the operators have continuous indication of the instrument's

availability. Thus, the amendment does not affect the margin of safety.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,

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

NRC Project Director: Theodore R. Quay.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 11, 1994.

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

3.1.f, ``Minimum Conditions for Criticality,'' in preparation for the

licensee's extension of the fuel cycle from 12 to 18 months. The

proposed amendment would revise TS 3.1.f to specify that the moderator

temperature coefficient (MTC) shall be no greater than 5.0 pcm/ deg.F

when at or below 60% rated thermal power and shall be zero or negative

when above 60% rated thermal power. The proposed amendment also

incorporates required actions to be implemented if the MTC

specification is not met.

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:

Significant Hazards Determination for Proposed Changes to TS 3.1.f

This change is being proposed in accordance with the provisions

of 10 CFR 50.92 to show that no significant hazards exist. The

proposed change will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Moderator temperature coefficient (MTC) is a physical

characteristic of the reactor core which affects the reactor's

response to transients. The MTC is not an accident initiator,

therefore changing the allowed MTC to 5.0 pcm/ deg.F at or below 60%

power, does not increase the probability of an accident previously

evaluated.

To verify that an increased MTC would not increase the

consequences of a previously evaluated accident, Wisconsin Public

Service Corporation (WPSC) reanalyzed the bounding MTC-related

events found in Kewaunee's Updated Safety Analysis Report (USAR).

Those events are:

1. Uncontrolled Rod Withdrawal at Intermediate Power (USAR

Section 14.1.2)

2. Rod Ejection at Hot Zero Power, Beginning of Cycle (USAR

Section 14.2.6)

3. Loss of External Electrical Load, Beginning of Cycle (USAR

Section 14.1.9)

4. Uncontrolled Boron Dilution (USAR Section 14.1.4)

5. Loss of Reactor Coolant Flow, Locked Rotor (USAR Section

14.1.8)

6. Loss of Reactor Coolant Flow, Both Pumps Trip (USAR Section

14.1.8)

Events numbered 1, 3, 4, 5, and 6 above were reanalyzed to

verify that the departure from nuclear boiling ratio (DNBR) would

not fall below the acceptance criterion of 1.300. In all five cases,

an assumed MTC value of 5.0 pcm/ deg.F at or below 60% power

increased the calculated DNBR value from the current USAR analyses.

Therefore the existing analyses associated with DNBR remain

bounding.

Event number 2 above was reanalyzed to ensure Kewaunee's peak

clad temperature (PCT) would not exceed the required acceptance

criterion of 2700 deg.F. When this accident was reanalyzed using the

positive MTC value of 5.0 pcm/ deg.F, the calculated PCT exceeded

2700 deg.F. Therefore it was necessary to decrease the allowable hot

channel factor (FQ) from 9.0 to 8.2. This change in allowable FQ

reduces a bounding value for future Kewaunee core designs and does

not have an adverse effect on plant safety limits or settings. This

FQ limit reduction for the accident of Rod Ejection at Hot Zero

Power, Beginning of Cycle, will be incorporated into the WPSC Reload

Safety Methodology and Kewaunee's USAR. The results of the analysis

using a FQ of 8.2 and a MTC value of 5.0 pcm/ deg.F indicate a lower

PCT than the previously bounding case.

In addition, WPSC reviewed the effects of this change on

Kewaunee's anticipated transient without scram (ATWS) analysis. This

review demonstrated that the ATWS analysis was found to include

assumptions bounding Kewaunee's proposed MTC limits.

The reanalyses of the six most limiting MTC-related transients

and review of the ATWS event demonstrate that this proposed change

does not involve an increase in the potential consequences of any

accidents previously evaluated.

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

accident from any accident previously evaluated.

The MTC is a physical characteristic of the reactor core and is

not an accident initiator. A changed MTC changes the reactor's

response to the postulated transients in chapter 14 of Kewaunee's

USAR. Therefore an MTC of 5.0 pcm/ deg.F at or below 60% rated

thermal power does not create the possibility of a new or different

type of accident from any accident previously evaluated.

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

The six most limiting MTC-related transients in chapter 14 of

the KNPP USAR were reanalyzed. Five of the six were reanalyzed to

verify the DNBR would not go below the DNBR acceptance criterion of

1.300. The results of these five DNBR-related analyses demonstrate

that the reanalyzed DNBR value is greater than the DNBR from the

currently bounding safety analyses. The sixth transient was

reanalyzed to verify the PCT limit of 2700 deg.F would not be

exceeded. Reanalysis of this sixth accident, Rod Ejection at Hot

Zero Power and Beginning of Cycle, indicated a need to lower the

allowable hot channel factor (FQ). Reanalysis incorporating this

more restrictive FQ value resulted in a calculated PCT value of

2504 deg.F, which is less than the previous PCT value of 2585 deg.F.

The new FQ limit will be incorporated into the next USAR revision,

which will be submitted in accordance with 10 CFR 50.71.

In addition, WPSC reviewed the effects of this change on

Kewaunee's anticipated transient without scram (ATWS) analysis. This

review demonstrated that the ATWS analysis was found to include

assumptions bounding Kewaunee's proposed MTC limits.

The reanalyses of the six most bounding transients and review of

the ATWS event demonstrate that restricting the MTC value to 5.0

pcm/ deg.F at or below 60% power does not reduce 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: University of Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: John N. Hannon.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: May 20, 1994.

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

3.4, ``Steam and Power Conversion System,'' by modifying and clarifying

the operability requirements for the main steam safety valves (MSSVs),

auxiliary feedwater system (AFW), and the condensate storage tank

system. The proposed amendment would eliminate inconsistencies within

Technical Specification Section 3.4 and would provide the basis for

operation of the AFW system below 15% reactor power.

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:

Significant Hazards Determination for Proposed Changes to Technical

Specification (TS) 3.4.a ``Main Steam Safety Valves''

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Currently, TS 3.4.a.1.A.2 requires five MSSV's to be operable

prior to heating the reactor > 350 deg.F. The proposed change

requires a minimum of two MSSVs per steam generator to be operable

prior to heating the reactor coolant system > 350 deg.F, and five

MSSVs per steam generator to be operable prior to reactor

criticality. If these conditions cannot be met within 48 hours,

within 1 hour action shall be initiated to achieve hot standby

within 6 hours, achieve hot shutdown within the following 6 hours,

and achieve and maintain the reactor coolant system temperature TM clad fuel, and to remove unnecessary detailed

descriptions of fuel and control rod assemblies.

Date of issuance: September 8, 1994.

Effective date: September 8, 1994.

Amendment Nos.: 110 and 101.

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

the Technical Specifications.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12366).

The August 4, 1994, submittal reformatted Section 5.6.1 and

restored and relocated to Section 5.6.1 the maximum enrichment limits

deleted in the February 16, 1994 amendment request, but did not change

the no significant hazards consideration as published in the Federal

Register.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 8, 1994.

No significant hazards consideration comments received: No

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

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

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

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

Date of application for amendments: September 28, 1993 (TS 93-10).

Brief description of amendments: The amendments clarify the

operability requirements for the fire suppression system flow path and

incorporate additional guidance into an action statement requirement

for spray and/or sprinkler systems inside containment.

Date of issuance: September 13, 1994.

Effective date: September 13, 1994.

Amendment Nos.: 186 and 178.

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

revise the technical specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59757). The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated September 13, 1994.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

Callaway County, Missouri

Date of application for amendment: January 27, 1993, clarified on

April 20, 1994.

Brief description of amendment: The amendment revises the Technical

Specification 6.3.1.2 to allow either the Health Physics Superintendent

or the Health Physics, Operations Supervisor to be designated as

Radiation Protection Manager (RPM).

Date of issuance: September 6, 1994.

Effective date: September 6, 1994.

Amendment No.: 92.

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

Technical Specification 6.3.1.2.

Date of initial notice in Federal Register: April 14, 1993 (58 FR

19490).

The clarifying information did not change the initial proposed no

significant hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 6, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-

339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of application for amendments: March 30, 1994.

Brief description of amendments: The amendments revise the NA-1&2

TS High Head Safety Injection (HHSI) surveillance requirements by

removing explicit numerical values and replacing them with broader non-

numerical requirements.

Date of issuance: September 6, 1994.

Effective date: September 6, 1994.

Amendment Nos.: 188 and 169.

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

the Technical Specifications.

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

22017)

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 6, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

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

and regulations. The Commission has made appropriate findings as

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

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

Because of exigent or emergency circumstances associated with the

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to prov

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