Biweekly Notice

UNITED STATES 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 May 27, 1994, through June 10, 1994. The

last biweekly notice was published on June 8, 1994 (59 FR 29623).

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, 11555 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 July 22, 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 intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

Branch, or may be delivered to the Commission's Public 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)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC 20555, and at the local public document

room for the particular facility involved.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: April 26, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications to change the Table 3.5-1 High

Containment Pressure ( Hi Level), Safety Injection Setting Limit from

less than or equal to 2.0 psig to less than or equal to 5.0 psig.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the High Containment Pressure (Hi Level)

actuation setting of [less than or equal to] 2.0 psig be revised to

[less than or equal to] 5.0 psig. This additional operating

flexibility will decrease the frequency of Containment venting

necessary to relieve containment of

non-condensible gases which build up during normal operation.

Based upon a statistical analysis of the containment pressure

channel uncertainty for a 30 month operating cycle, a margin must be

allowed between the Technical Specification limit (plant setting)

and the Safety Analysis limit so that the Safety Analysis limit(s)

will not be exceeded under the worst circumstances. For a Technical

Specification value of [less than or equal to] 5.0 psig, the

corresponding Safety Analysis limit must be increased to 10 psig to

provide margin for the channel statistical allowance. A safety

evaluation performed pursuant to 10 CFR 50.59 is on file which

supports a change in the Safety Analysis limit from 7.3 psig

(current value) to 10.0 psig. Key conclusions of the Safety

Evaluation are that neither the probability nor the consequences of

an accident or malfunction of equipment important to safety

previously evaluated in the Safety Analysis report would be

increased.

Thus, assurance is provided that appropriate protective actions

in accordance with the Technical Specifications will be taken so

that Safety Analysis limits are not exceeded.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in the Technical Specification limit

together with the change in the Safety Analysis limit provides

adequate margin to accommodate instrument channel uncertainty over a

30 month operating cycle. Plant equipment, which would be set at the

Technical Specification limit, will therefore provide protective

functions to assure that safety analysis limits are not exceeded.

This would prevent the possibility of a new or different kind of

accident from that previously evaluated from occurring.

3. There has been no reduction in the margin of safety.

The proposed change to the Technical Specification limit would

decrease the frequency of containment purges necessary to vent the

build up of non-condensible gases during normal operation. This

would result in a decrease in the amount of radioactivity discharged

to the environment (due to decay), decrease the potential for high

Containment pressure alarms and increase the margin for an ESF trip.

The change to the Safety Analysis limits, justified by a safety

Evaluation performed in accordance with 10 CFR 50.59, assures

sufficient margin exists to accommodate channel instrument

uncertainty over the maximum operating cycle length. This margin is

necessary so that safety functions will occur and Safety Analysis

limits will be preserved.

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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: Michael L. Boyle

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: May 24, 1994

Description of amendment request: The proposed amendments would

transfer the boron concentration in Technical Specification (TS) 3.9.1

for the reactor coolant system and the refueling canal during MODE 6,

and the boron concentration in TS 3.9.12 for the spent fuel pool from

the TS to the Core Operating Limits Report (COLR). The application is

submitted in response to the guidance in Generic Letter 88-16 which

addresses the transfer of fuel cycle-specific parameter limits from the

TS to the 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:

The following analysis, performed pursuant to 10 CFR 50.91,

shows that the proposed amendment will not create a significant

hazards consideration as defined by the criteria of 10 CFR 50.92.

1. This amendment will not significantly increase the

probability or consequence of any accident previously evaluated.

No component modification, system realignment, or change in

operating procedure will occur which could affect the probability of

any accident or transient. The relocation of boron concentration

values to the COLR is an adminsitrative change which will have no

effect on the probability or consequences of any previously-analyzed

accident. The required values of boron concentration will continue

to be determined through use of approved methodologies.

2. This amendment will not create the possibility of any new or

different accidents not previously evaluated.

No component modification or system realignment will occur which

could create the possibility of a new event not previously

considered. The administrative change of relocating parameters to

the COLR, in this case boron concentration, cannot create the

probability of an accident.

3.This amendment will not involve a significant reduction in a

margin of safety.

Required boron concentrations will remain appropriate for each

cycle, and will continue to be calculated using approved

methodologies. There is no significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: David B. Matthews, Director

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: May 12, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification Sections 3.1 and 4.1 for Protective

Instrumentation, the associated bases, and tables to increase the

surveillance test intervals (STIs) and add allowable out-of-service

times (AOTs). All proposed STI and AOT changes are in accordance with

General Electric Company Licensing Topical Reports (LTRs) which have

been previously reviewed and approved by the NRC staff. Also, AOTs are

clarified in accordance with the most recently approved BWR Owners'

Group letters which were used in the development of NUREG-1433

``Standard Technical Specifications, General Electric Plants, BWR/4.''

The Technical Specification changes will permit specified Channel Tests

to be conducted quarterly rather than weekly or monthly. The amendment

will enhance operational safety by reducing 1) the potential for

inadvertent plant scrams, 2) excessive test cycles on equipment, and 3)

the diversion of plant personnel and resources on unnecessary testing.

Two additional technical changes are proposed. The first change

involves extending the Channel Calibration interval for average power

range monitor (APRM) scram instrumentation from weekly to quarterly.

GPUN has evaluated the effect of drift on the setpoint over the longer

interval for this instrumentation and found it to be acceptable. The

second change would add a quarterly Channel Calibration requirement for

High Drywell Pressure (for Core Cooling) and Turbine Trip Scram

instrumentation. This would be a new requirement not currently

incorporated in the Technical Specifications.

Nineteen editorial changes have been incorporated in

Instrumentation Sections 3.1 and 4.1 to provide clarity and

consistency. These items are editorial only and do not alter the

meaning or intent of any requirements. Examples of editorial changes

are: 1) capitalize definitions where used, 2) punctuation and

grammatical corrections, 3) ensuring consistency in STI nomenclature,

and 4) reformat of tables. A table note and its associated footnote

were deleted which involved a 1985 licensing condition which is no

longer applicable.

Basis for proposed no significant hazards consideration

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

provided its analyses of the issue of no significant hazards

consideration, which is presented below:

NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION OF TECHNICAL

CHANGES

1. The operation of the Oyster Creek Nuclear Generating Station,

in accordance with the proposed amendment, will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The generic analysis contained in LTR NEDC-30851P-A assessed the

impact of changing RPS [reactor protection system] STIs and adding

AOTs on RPS failure frequency, scram frequency and equipment

cycling. Specifically, Section 5.7.4, ``Significant Hazards

Assessment,'' of NEDC-30851P-A states that:

Fewer challenges to the safeguards system, due to less frequent

testing of the RPS, conservatively results in a decrease of

approximately one percent in core damage frequency. This decrease is

based upon the following:

Based on the plant-specific experience presented in Appendix J,

the estimated reduction in scram frequency (0.3 scrams/yr)

represents a 1 to 2 percent decrease in core damage frequency based

on the BWR plant-specific Probabilistic Risk Assessments (PRAs)

listed in Table 5-8.

The increase in core damage frequency due to less frequent

testing is less than one percent. This increase is even lower (less

than 0.01 percent) when the changes resulting from the

implementation of the Anticipated Transients Without Scram (ATWS)

rule are considered. Therefore, this increase is more than offset by

the decrease in CDF [core damage frequency] due to fewer scrams.

The effect of reducing unnecessary cycles on RPS equipment,

although not easily quantifiable also results in a decrease in core

damage frequency.

The overall impact on core damage frequency of the changes in

allowable out-of-service times is negligible.

The BWR Owners' Group concluded that the proposed changes do not

significantly increase the probability or consequences of an

accident previously evaluated since the increase in probability of a

scram failure due to RPS unavailability is insignificant. The

overall probability of an accident is decreased as the time RPS

logic operates as designed is increased resulting in less

inadvertent scrams during testing and repair. The plant-specific

evaluation performed by GPUN and GE demonstrates that while the

Oyster Creek RPS differs from the generic model analyzed in the RPS

LTR (NEDC-30851P-A), the net effect of the differences do not alter

the generic conclusions. The AOTs proposed for RPS instrumentation

are based on improved wording developed for use in NUREG 1433,

``Standard Technical Specifications, General Electric Plants, BWR/

4,'' which ensures a loss of function does not occur. In addition,

the change to the APRM Scram Channel Calibration surveillance

interval from weekly to quarterly has been evaluated by GPUN to

determine the effect on setpoint drift. The results of the

evaluation show acceptable performance of this scram parameter

ensuring that the safety analysis remains valid. The clarification

that a Channel Calibration is not applicable to Turbine Trip Scram

instrumentation is appropriate since this trip parameter senses

turbine stop valve position via limit switches which are fixed in

position and adjusted, as necessary, during valve maintenance. This

trip parameter and its switch adjustment methods are similar to the

Main Steamline Isolation Valve [MSIV] Scram for which the Technical

Specifications require only a Channel Test.

LTR NEDC-30936P-A (Parts 1 and 2) contains an assessment of the

impact of changing STIs and AOTs for BWR ECCS Actuation

Instrumentation. Section 4.0, ``Technical Assessment of Changes,''

of NEDC-30963P-A (Part 2) states that:

The results indicate an insignificant (less than 5E-7 per year)

increase in water injection function failure frequency when STIs are

increased from 31 days to 92 days, AOTs for repair of the ECCS

actuation instrumentation are increased from one hour to 24 hours,

and AOTs for surveillance testing are increased from two to six

hours. For all four BWR models the increase represents less than 4%

increase in failure frequency. However, when other factors which

influence the overall plant safety are considered, the net result is

judged to be an improvement in plant safety.

From this generic analysis, the BWR Owners' Group concluded that

the proposed changes do not significantly increase the probability

or consequences of an accident previously evaluated since the

increase in probability of a water injection failure due to ECCS

instrumentation unavailability is insignificant and the net result

is judged to be an improvement in plant safety. The plant-specific

evaluation performed by GPUN and GE demonstrates that while the

Oyster Creek ECCS differs from the generic model analyzed in LTR

NEDC-30936P-A, the net effect of the plant-specific differences do

not alter the generic conclusions. The addition of a quarterly

Channel Calibration STI for the High Drywell Pressure ECCS

initiation parameter is consistent with the calibration interval

requirement for other similar instrumentation at Oyster Creek and

ensures the regular performance of calibrations. This is a new

requirement not currently contained in the Technical Specifications

and experience performing the High Drywell Pressure (Core Cooling)

instrument calibration at a quarterly interval has proven adequate

for instrument performance monitoring.

LTRs NEDC-30851P-A, Supplement 2 and NEDC-31677P-A contain

generic analyses assessing the impact of changing STIs and AOTs for

BWR Isolation Actuation Instrumentation which are common or not

common to RPS and ECCS instrumentation. Section 4.0, ``Summary of

Results,'' of NEDC-30851P-A, Supplement 2 states that:

The results indicated that the effects on probability of failure

to initiate isolation are very small and the effects on probability

or frequency of failure to isolate are negligible in nearly every

case. In addition, the results indicated that increasing the AOT to

24 hours for tests and repairs has a negligible effect on the

probability of failure of the isolation function. These combined

with changes to the testing intervals and allowed out-of-service

times for RPS and ECCS instrumentation provide a net improvement to

plant safety and operations.

and Section 5.6, ``Assessment of Net Effect of Changes,'' of

NEDC-31677P-A states that:

A reduction in core damage frequency (CDF) of at least as much

as estimated in the ECCS instrumentation analysis can be expected

when the isolation actuation instrumentation STIs are changed from

one month to three months. The chief contributor to this reduction

is the channel functional tests for the MSIVs. Inadvertent closure

of the MSIVs will cause an unnecessary plant scram. This reduction

in CDF more than compensates for any small incremental increase (10%

or 1.0E-07/year) in calculated isolation function failure frequency

when the STI is extended to three months.

Based on this generic analysis, the BWR Owners' Group concluded

that the proposed changes do not significantly increase the

consequences of an accident previously evaluated since the increase

in probability of an isolation failure due to isolation

instrumentation unavailability is insignificant. The proposed

wording of the AOTs is based on the clarifications used in the

development of NUREG 1433, ``Standard Technical Specifications,

General Electric Plants, BWR/4,'' which ensures a loss of function

does not occur where applied to isolation actuation instrumentation.

LTR NEDC-30851P-A, Supplement 1 contains a generic analysis

assessing the impact of changing control rod block STIs on Rod Block

failure frequency. Section 5 (Brookhaven National Laboratory

Technical Evaluation Report - Attachment 2 to the NRC SER) of NEDC-

30851P-A, Supplement 1 states that:

The BWROG proposed changes to the Technical Specifications

concerning the test requirements for BWR control rod block

instrumentation. The changes consist of increasing the surveillance

test intervals form one to three months. These test interval

extensions are consistent with the already approved changes to STIs

for the reactor protection system. The technical analysis reviewed

and verified as documented herein indicates that there will be no

significant changes in the availability of the control rod block

function if these changes are implemented. In addition, there will

be a negligible impact on the plant core melt frequency due to the

decreased testing.

Bases contained in GE Topical Report GENE-770-1-A assessed the

impact of changing STIs and AOTs on failure frequency for selected

systems. Section 2.0, ``Summary,'' of GENE-770-06-1-A states that:

Technical bases are provided for selected proposed changes to

the instrumentation STIs and AOTs that were identified in the BWROG

Improved BWR Technical Specification activity. These STI and AOT

changes are consistent with approved changes to the RPS, ECCS, and

isolation actuation instrumentation. These proposed changes do not

result in a degradation to overall plant safety.

The BWR Owners' Group concluded from the generic analysis in

NEDC-30851P-A, Supplement 1 and the bases in GENE-770-06-1-A that

the proposed changes do not significantly increase the probability

or consequences of an accident previously evaluated. GPUN's

utilization of GENE-770-06-1-A is limited to the identified AOTs for

Control Rod Block instrumentation analyzed in NEDC-30851P-A since

the Control Rod Block LTR did not explicitly address AOTs.

2. The operation of Oyster Creek Nuclear Generating Station, 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 addition of allowable out-of-service times (AOTs) consistent

with wording developed for use in Improved Standard Technical

Specifications to ensure no loss of function and the revision of

surveillance test intervals (STIs) does not alter the function of

RPS, ECCS, Isolation or Rod Block instrumentation nor involve any

type of plant modification. No new modes of plant operation are

involved with the changes.

Adding a quarterly Channel Calibration STI for High Drywell

Pressure instrumentation (for Core Cooling) establishes a

requirement in the Technical Specifications which is not currently

incorporated. This is an additional requirement beyond that already

in place for this instrumentation and will not alter its operation

since by their nature STIs ensure proper instrument performance. The

clarification that a Channel Calibration is not applicable to

Turbine Trip Scram instrumentation is appropriate since this trip

parameter senses turbine stop valve position via limit switches

which are fixed in position and adjusted during valve maintenance.

This trip parameter and its switch adjustment methods are similar to

the Main Steamline Isolation Valve Scram for which the Technical

Specifications require only a Channel Test. Revising the Channel

Calibration STI for APRM Scram instruments from weekly to quarterly

allows these instruments to benefit from the Channel Test STI change

provided by the generic analysis in the RPS LTR. The benefits

include a significant reduction in the number of half-scram states

the plant will undergo reducing the potential for inadvertent plant

trips. The effect of setpoint drift over the longer interval has

been evaluated and found acceptable.

The proposed changes will not alter the physical characteristics

of any plant systems or components and all safety-related systems

and components remain within their applicable design limits. Thus,

system and component performance is not adversely affected by these

changes, thereby assuring that the design capabilities of those

systems and components are not challenged in a manner not previously

assessed so as to create the possibility of a new or different kind

of accident.

3. The operation of the Oyster Creek Nuclear Generating Station,

in accordance with the proposed amendment, will not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed and approved the generic studies

contained in the GE Licensing Topical Reports and has concurred with

the BWR Owners' Group that the proposed changes do not significantly

affect the availability of RPS, ECCS Actuation, Isolation Actuation

and Control Rod Block instrumentation. The proposed addition of

allowable out-of-service times for instruments addressed by the LTRs

provides reasonable times for making repairs and performing tests.

The lack of sufficient out-of-service time provided in current

Technical Specifications, increases the potential for an inadvertent

scram or equipment actuation. The proposed AOTs provide realistic

times to complete required actions without increasing overall

instrument failure frequency and ensure that no loss of function

occurs, therefore, there is no significant reduction in the margin

of safety.

The LTRs demonstrate that extending surveillance test intervals

does not result in significant changes in the probability of

instrument failure. Where Channel Calibration frequency has not

changed, assurance exists that setpoints will not be affected by

drift. In the case of the APRM Scram Channel Calibration, the

proposed change to quarterly from weekly has been evaluated and

found acceptable. Expected instrument performance over the extended

interval will assure that applicable safety analyses will continue

to be met. In addition, other instrumentation was evaluated for

drift effects of setpoints and was found acceptable. The addition of

a quarterly Channel Calibration interval for High Drywell Pressure

(for Core Cooling) is consistent with Channel Calibration STIs for

most other instrumentation at Oyster Creek and has been the interval

used to achieve an adequate level of instrument performance

monitoring. The clarification that a Channel Calibration is not

applicable to Turbine Trip Scram instrumentation ensures consistency

in the establishment of surveillance requirements. This trip

parameter senses turbine stop valve position via limit switches

which are fixed in position and adjusted during valve maintenance.

This trip parameter and its switch adjustment methods are similar to

the Main Steamline Isolation Valve Scram for which the Technical

Specifications require only a Channel Test. These proposed changes,

when coupled with the reduced probability of test-induced plant

transients and equipment failures, do not result in a reduction in

the margin of safety.

No Significant Hazards Consideration Evaluation For Editorial

Changes

The above nineteen proposed changes are editorial in nature and

are typical example I.c.2.e.i in 51FR7744. Therefore, they do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The editorial changes described above do not change the design

or operation of any structure, system or component relied upon to

prevent or mitigate the consequences of any accident evaluated.

These editorial changes also do not add new structures, systems or

components which may have an effect on existing elements of the

facility. The changes proposed correct, clarify and/or retain

existing requirements.

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

form any accident previously evaluated.

Since neither physical changes to the facility nor changes in

its operation are involved in the proposed editorial changes to the

Technical Specifications, there is no possibility for creation of a

new or different kind of accident.

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

Facility configuration and operation are unaffected by the

proposed editorial changes. As a result no changes in margin of

safety occur.

The editorial changes described and evaluated above are purely

administrative to achieve consistency or correct an error in the

Technical Specifications.

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

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

satisfied for both the technical issues and editorial changes.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753

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

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

NRC Project Director: John F. Stolz

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

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

Date of amendment request: February 10, 1994

Description of amendment request: The revision proposed by

Technical Specification Change Request (TSCR) No. 230 to the Technical

Specifications would revise specification 3.7.2.c, ``Unit Electric

Power System,'' to eliminate testing of an emergency diesel generator

(EDG) when the redundant EDG is inoperable.

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 serves to assure that an EDG is always

available to perform on demand and the lower number of demands for

performance reduce the probability of equipment failure. The

required action no longer requires a ``test'' be performed.

Therefore, the word ``test'' has been deleted from TS 3.7.2.c. The

change is administrative. Since the proposed amendment does not

affect the design or performance of the diesel generators or their

ability to perform their design function, the change will not result

in an increase in the consequences or probability of an accident

previously analyzed. The proposed change will increase diesel

generator reliability, thereby increasing overall plant safety.

2. Operation of the facility in accordance with the proposed

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

of accident from any accident previously evaluated. Accidents

involving loss of off-site power and single failure have been

previously evaluated. The change does not introduce any new mode of

plant operation or new accident precursors, involve any physical

alterations to plant configurations, or make any changes to system

setpoints which could initiate a new or different kind of accident.

3. Operation of the facility in accordance with the proposed

amendment does not involve a significant reduction in a margin of

safety. This change does not result in a reduction in the margin of

safety since there is no margin of safety associated with the

supplemental immediate and daily testing of the operable EDG. If a

margin of safety were presumed to exist, no reduction would result

because of the proposed amendment: no physical modification to the

plant or change to procedurally prescribed operator actions resulted

from the proposed amendment.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz

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

YankeeAtomic Power Station, Lincoln County, Maine

Date of amendment request: May 25, 1994

Description of amendment request: The proposed amendment would 1)

allow entry through an operable personnel air lock hatch to perform

surveillance testing, repair an inoperable hatch, or perform other

necessary activities inside containment, 2) update plant Technical

Specifications to reflect a previous change to the list of containment

boundary valves, 3) add a new exception to allow quarterly surveillance

testing of the excess flow check valves, 4) add a new exception to

allow periodic preventive maintenance on control room ventilation

lasting up to 30 minutes per calendar quarter without a written report

of such inoperability, and 5) make related administrative changes to

reflect and clarify items 1 through 4 above.

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 staff's analysis is

presented below:

1. The proposed amendment would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Containment air lock hatch entry, surveillance testing of the

excess flow check valves, and preventive maintenance of control room

ventilation are of short duration and do not alter any associated

remedial action completion times, or the requirements of Technical

Specification 3.0.A. If necessary, prompt operator action to restore

containment integrity, excess flow check valve position, or control

room ventilation is assured by plant operators, or individual(s)

procedurally dedicated to perform such restoration. The subject

containment boundary valves are manual containment isolation valves,

and the current specification allows them to be repositioned under

administrative control without compensatory measures to isolate the

penetration. The boundary valves to be added remain closed during

power operation, and are opened only after the reactor is shut down

and cooldown has begun. The boundary valves to be deleted are open

only during plant heatup.

The staff therefore concludes that implementation of the proposed

change will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change to containment air lock hatch entry,

surveillance testing of the excess flow check valves, and preventive

maintenance of the control room ventilation system, will not affect

equipment reliability when such equipment is required to be

operable. The Limiting Conditions for Operation and Remedial Actions

for these items remain unchanged to govern operability of the

equipment. The containment boundary valves being added are closed

when the reactor is at power, and are opened only after the reactor

is shut down. The boundary valves being deleted are open only during

plant heatup. The subject boundary valves are manual containment

isolation valves, and the current specification allows them to be

repositioned under administrative control without compensatory

measures to isolate the penetration.

The staff therefore concludes that implementation of the

proposed change will not create any 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.

The proposed change would allow excess flow check valves to be

exercised through approximately 1.5 inches of valve travel on a

quarterly basis without declaring the valves inoperable or taking

compensatory measures. Such testing constitutes approximately 15

minutes per calendar quarter, during which time containment

isolation can easily be reestablished. Similarly, access through an

operable air lock hatch would allow the hatch to be open for only a

short period of time and while under control of an individual

dedicated to operating the hatch. The proposed change also permits

the control room ventilation system to be inoperable for 30 minutes

per calendar quarter, without a written report of such

inoperability. Because of the short time during which these systems

are unavailable, and because operation is easily reestablished,

there is no significant reduction in a margin of safety. The

containment boundary valves being added are closed when the reactor

is at power, and are opened only after the reactor is shut down. The

boundary valves being deleted are open only during plant heatup. The

subject boundary valves are manual containment isolation valves, and

the current specification allows them to be repositioned under

administrative control without compensatory measures to isolate the

penetration.

The staff therefore concludes that implementation of the proposed

change would not involve a significant reduction in a 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: Wiscasset Public Library, High

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

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 83 Edison Drive, Augusta, Maine 04336

NRC Project Director: Walter R. Butler

Northeast Nuclear Energy Company (NNECO), Docket No. 50-

245,Millstone Nuclear Power Station, Unit 1, New London County,

Connecticut

Date of amendment request: April 29, 1994

Description of amendment request: The amendment would change the

requirement for reactor operators (RO) in Table 6.2-1 from 2 to 3 for

the RUN, STARTUP/HOT STANDBY and HOT SHUTDOWN conditions. In addition,

two typographical corrections are made to page 6-4.

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:

NNECO has reviewed the proposed changes in accordance with

10CFR50.92 and concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed changes do not involve a significant

hazards consideration because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Accident analyses for Millstone Unit No. 1 do not require a

specific number of operators. Increasing the Technical Specification

minimum to require a third RO does not decrease the effectiveness of

the shift staff in response to normal or abnormal conditions. In

fact, the third RO enhances the ability of the operating crew to

mitigate complex transients which could occur during beyond design

basis events. The shifts have trained and functioned at the higher

staffing level for several years.

The typographical corrections to page 6-4 provide a clearer

representation of the required actions, and do not affect the intent

nor implementation of the specification.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of a previously analyzed

accident.

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

from any previously analyzed.

The addition of a third RO to the minimum shift-crew composition

required by Technical Specification Table 6.2-1 does not affect the

operation of the unit, nor does it change any of the operating

procedures, off-normal procedures, or EOPs [emergency operating

procedures]. Staffing the control room with an additional operator

enhances the capability of the operating crew to mitigate

transients. Therefore, addition of a third RO to the minimum shift-

crew composition cannot create the possibility of a new or different

accident.

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

The proposed addition of a third reactor operator is to ensure

that sufficient operating staff is available to respond to complex

transients involving multiple equipment failures. Ensuring that

sufficient resources are available to cope with beyond design basis

event scenarios provides an increase in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

London Turnpike, Norwich, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz

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: May 6, 1994

Description of amendment request: The proposed amendment would

modify the Limiting Conditions for Operation (LCO) for the Millstone

Unit 2 Technical Specifications 3.8.2.3 and 3.8.2.4 and the

surveillance requirement of TS 4.8.2.3.2.c.3. These changes relate to

the amperage requirements and the charging capability of the DC

distribution systems.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

In 1993, revised battery and battery charger sizing calculations

demonstrated that a charger capacity of 400 amperes is sufficient to

provide the continuous DC loads, and is capable of recharging a

fully discharged station battery in a timely manner consistent with

the design basis discussed in Section 8.5.3.1 of the Millstone Unit

No. 2 FSAR [Final Safety Analysis Report]. The calculations

determined that the largest continuous load was 154 amperes;

therefore, 400 amperes of charging capacity could provide 246

amperes to recharge a battery.

The calculations conservatively demonstrated that this charging

capacity could recharge a battery in 10.37 hours. This recharging

time is well within the 12-hour recharging time discussed in Section

8.5.3.1 of the Millstone Unit No. 2 FSAR. Additionally, this

recharging time is more conservative than the 24-hour recharging

time stated in Section 8.3.2 of the original Safety Evaluation for

Millstone Unit No. 2. Therefore, the proposed changes do not involve

a significant increase in the probability or consequences of an

accident previously analyzed.

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

from any previously analyzed.

The proposed LCO and surveillance changes do not alter the

existing DC bus configuration, as described in Section 8.5.3.1 of

the Millstone Unit No. 2 FSAR. This bus configuration has been

previously analyzed, and was found acceptable. The proposed changes

also meet the recharging time specified in the design basis.

Therefore, the proposed changes do 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.

In 1993, revised battery and battery charger sizing calculations

demonstrated that a charger capacity of 400 amperes is sufficient to

provide the continuous DC loads, and is capable of recharging a

fully discharged station battery in a timely manner consistent with

the design basis discussed in Section 8.5.3.1 of the Millstone Unit

No. 2 FSAR. The calculations determined that the largest continuous

load was 154 amperes; therefore, 400 amperes of charging capacity

could provide 246 amperes to recharge a battery. The calculations

conservatively demonstrated that this charging capacity could

recharge a battery in 10.37 hours. This recharging time is well

within the 12-hour recharging time discussed in Section 8.5.3.1 of

the Millstone Unit No. 2 FSAR. Additionally, this recharging time is

more conservative than the 24-hour recharging time stated in Section

8.3.2 of the original Safety Evaluation for Millstone Unit No. 2.

Therefore, the proposed changes do 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, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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: May 6, 1994

Description of amendment request: The proposed amendment would

provide additional Technical Specification requirements regarding non-

Quality Assurance (QA) equipment utilized to achieve feedwater

isolation in response to a main steam line break (MSLB) inside

containment. Specifically the amendment would incorporate additional

sections numbered 3/4.7.1.6, titled ``Plant Systems - Main Feedwater

Isolation Components (MFICs);'' 3/4.8.2.1A, titled '' Onsite Power

Distribution Systems - A.C. Distribution - Operating;'' and 3/4.8.2.5,

titled ``Onsite Power Distribution Systems (Turbine Battery) - D.C.

Distribution - Operating.'' In addition, the proposed amendment would

modify the Index and the Bases to reflect the additional requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed changes in accordance with

10CFR50.90 and has concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Currently, the Millstone Unit No. 2 Technical Specifications

contain response time requirements for the feedwater isolation

valves to ensure rapid isolation of feedwater to the steam

generators and to maintain the peak containment pressure below the

containment design pressure of 54 psig. However, clear Action

Statements specifying operability requirements for the non-QA

equipment associated with feedwater isolation are not included

within the Millstone Unit No. 2 Technical Specifications. NNECO's

proposal to add sections 3/4.7.1.6, 3/4.8.2.1A, and 3/4.8.2.5 into

the Millstone Unit No. 2 Technical Specifications will incorporate

additional requirements regarding components that are credited to

provide feedwater isolation in the event of an MSLB inside

containment. These proposed changes will impose additional

limitations, restrictions, and controls not currently in place in

the Millstone Unit No. 2 Technical Specifications.

Additionally, NNECO's proposals to modify the Bases and the

Index of the Millstone Unit No. 2 Technical Specifications will: 1)

provide personnel with information concerning the additional

requirements, and 2) correct an editorial error. These proposed

changes to the Bases and the Index do not alter the manner in which

equipment is operated, nor do they affect equipment availability.

Based on the above, the proposed license amendment does not

involve a significant increase in the probability or consequences of

an accident previously analyzed.

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

from any previously analyzed.

NNECO's proposal to add section 3/4.7.1.6, 3/4.8.2.1A, and 3/

4.8.2.5 into the Millstone Unit No. 2 Technical Specifications will

incorporate additional requirements regarding components that are

credited to provide feedwater isolation in the event of an MSLB

inside containment. These proposed changes will impose additional

limitations, restrictions, and controls not currently in place in

the Millstone Unit No. 2 Technical Specifications.

Additionally, NNECO's proposals to modify the Bases and the

Index of the Millstone Unit No. 2 Technical Specifications will: 1)

provide personnel with information concerning the additional

requirements, and 2) correct an editorial error. These proposed

changes to the Bases an the Index do not alter the manner in which

equipment is operated, nor do they affect equipment availability.

Based on the above, the proposed license amendment cannot create

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

previously analyzed.

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

NNECO's proposal to add sections 3/4.7.1.6, 3/4.8.2.1A, and 3/

4.8.2.5 into the Millstone Unit No. 2 Technical Specifications will

incorporate additional requirements regarding components that are

credited to provide feedwater isolation in the event of an MSLB

inside containment. These proposed changes will impose additional

limitations, restrictions, and controls not currently in place in

the Millstone Unit No. 2 Technical Specifications.

Additionally, NNECO's proposals to modify the Bases and the

Index of the Millstone Unit No. 2 Technical Specifications will: 1)

provide personnel with information concerning the additional

requirements, and 2) correct an editorial error. These proposed

changes to the Bases and the Index do not alter the manner in which

equipment is operated, nor do they affect equipment availability.

Therefore, this proposed license amendment does not involve a

significant reduction in a margin of safety. In fact. The margin of

safety will be increased due to the imposition of restriction on the

non-QA equipment credited for feedwater isolation in the event of an

MSLB inside containment.

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, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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

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

Connecticut

Date of amendment request: May 6, 1994

Description of amendment request: The proposed amendment modifies

the monthly operational test of the reactor trip bypass breakers to

monthly staggered, such that each breaker is tested every 62 days.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in probability or consequences

of an accident previously evaluated.

Revising the technical specifications to require a staggered

monthly surveillance operational test of the reactor trip bypass

breakers (such that each breaker is tested every 62 days) will only

make operational testing of the reactor trip bypass breakers

consistent with operational testing of the trip breakers and the

automatic trip and interlock logic. It will also reduce cycling of

the reactor trip bypass breakers by eliminating the requirement to

test both bypass breakers during the monthly surveillance, thereby

reducing maintenance and surveillance time. The proposed changes do

not affect any of the design basis accidents nor are there any

malfunctions associated with these changes.

Additionally, this technical specification bases change only

clarifies both the meaning of a reactor trip breaker and trip

breaker train which have been included for completeness and clarity

concerning the reactor trip breaker system.

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

previously evaluated.

Revising the technical specifications to require a staggered

monthly surveillance operational test of the reactor trip bypass

breakers (such that each breaker is tested every 62 days) will only

make operational testing of the reactor trip bypass breakers

consistent with operational testing of the reactor trip breakers and

the automatic trip and interlock logic. There are no new failure

modes associated with the proposed changes. Since the plant will

continue to operate as designed, the proposed changes will not

modify the plant response to the point where it can be considered a

new accident.

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

Revising the technical specifications to require a staggered

monthly surveillance operational test of the reactor trip bypass

breakers (such that each breaker is tested ever 62 days) will only

make operational testing of the reactor trip bypass breakers

consistent with operational testing of the reactor trip breakers and

the automatic trip and interlock logic. It will also reduce cycling

of the reactor trip bypass breakers by eliminating the requirement

to test both bypass breakers during the monthly surveillance,

thereby reducing maintenance and surveillance time. The proposed

changes do not have any adverse impact on the protective boundaries

nor do they affect the consequences of any accident previously

analyzed. The surveillance requirements will still ensure that the

reactor trip breakers and the reactor trip bypass breakers are

tested and within the limits. Therefore, the proposed changes will

not impact the margin of safety as designated in the bases of any

technical specification.

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

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

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

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, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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

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

Connecticut

Date of amendment request: May 18, 1994

Description of amendment request: The amendment would change

operability requirements for the Fuel Building Exhaust Filter System to

require it to be operable whenever irradiated fuel is in the spent fuel

pool, which has had less than 60 days of decay time. Surveillance

requirements for the Fuel Building Exhaust Filter System would be

changed to require that the system be tested and verified operable at

no greater than 31 days prior to its required usage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

...The proposed change does not involve an SHC [significant

hazards consideration] because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed modification will revise the period of time during

which the Fuel Building Exhaust Filter System must be operable.

The propose[d] change will require that the system is operable

whenever irradiated fuel, which has decayed less than 60 days, is in

the spent fuel pool. Currently, the system is required to be

operable whenever a load is moved over the pool or fuel is being

moved in the pool.

The modification has no effect on the probability of a fuel

handling accident. The consequences of a fuel handling accident has

been evaluated at two intervals. The first time is the minimum decay

time. At this time (t=100 hours) with irradiated fuel in the pool,

the Fuel Building Exhaust Filter System is required, per the

existing and the proposed Technical Specification, to be operable.

Therefore, the consequences of an accident are identical to that

described in the FSAR [Final Safety Analyses Report]. The second

scenario evaluated is when the filters are initially isolated (t=60

days). The resultant offsite dose, assuming no filtration and lower

core inventory due to decay, are significantly lower than was

calculated at t=100 hours. Therefore, the existing accident analysis

in FSAR Section 15.7.4 is limiting and the proposed modification

will not impact the probability or consequences of an accident.

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

from any previously evaluated.

The proposed change does not impact any system or component

which could cause a fuel handling accident. The Fuel Building

Exhaust Filter System is used for accident mitigations. It's failure

cannot, in any way, create the possibility of a new or different

kind of accident.

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

The proposed change to the Fuel Building Exhaust Filter System

has been analyzed at the two most critical times. The first analysis

was done when the fuel is first placed in the pool, and the second

analysis was done when the filtration system is isolated. The first

event resulted in no change in assumptions in the analysis presented

in the FSAR, ergo no change in dose. The second event has been

analyzed and doses have decreased, when compared to the first event.

The system will be verified operable per the performance of

Surveillance Requirement 4.9.12a prior to fuel or load movement over

the pool. Therefore, there is no 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, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: May 13, 1994

Description of amendment request: This amendment would revise

Technical Specifications Surveillance Requirement 4.8.1.1.2e.8, which

requires that an emergency diesel generator be retested within 5

minutes after completing a 24-hour endurance run.

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 Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS change would revise the Emergency Diesel

Generator (EDG) surveillance criteria to allow the hot restart test

to be performed independent of the Engineered Safety Features (ESF)

load sequencing test and the 24 hour endurance run. The proposed

surveillance requirements would continue to demonstrate that the

objectives of each of these tests are met. Specifically, the EDG's

are shown to be capable of starting the ESF loads in the required

sequence, operating at full load for an extended period of time, and

restarting from a full load temperature condition. Therefore, the

proposed changes would not adversely affect the EDG's ability to

support mitigation of the consequences of any previously evaluated

accident. The proposed changes to the surveillance requirements do

not affect the initiation or progression of any accident sequence.

Therefore, the proposed change does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

This proposed TS change does not require physical changes to the

plant or equipment, and does not impact any design or functional

requirements of the Emergency Diesel Generators (EDGs). The proposed

change affects surveillance test criteria such that increased

scheduling flexibility is allowed while the test objectives

associated with demonstrating EDG operability continue to be met.

The proposed changes do not allow any plant configurations that are

presently prohibited by the Technical Specifications.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The proposed TS change does not involve a change to the physical

design or functional requirements of the Emergency Diesel Generators

(EDGs). Surveillance testing in accordance with the proposed

Technical Specification will continue to demonstrate the ability of

the EDG's to perform their intended function of providing electrical

power to ESF systems needed to mitigate design basis transients,

consistent with the plant safety analyses. The margin of safety

demonstrated by the plant safety analyses is therefore not affected

by the proposed change.

Therefore, the proposed TS 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller

Philadelphia Electric Company, Public Service Electric and Gas

Company,Delmarva Power and Light Company, and Atlantic City

Electric Company,Docket No. 50-277, Peach Bottom Atomic Power

Station, Unit No. 2,York County, Pennsylvania

Date of application for amendment: May 13, 1994

Description of amendment request: The proposed amendment would

extend the Type A test (i.e., Containment Integrated Leak Rate Test)

interval on a one-time basis.

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 Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The accidents which are potentially negatively impacted by the

proposed change are any Loss of Coolant Accident (LOCA) inside

primary containment as described in the PBAPS [Peach Bottom Atomic

Power Station], Units 2 and 3 UFSAR [Updated Final Safety Analysis

Report].

The proposed change increases the surveillance interval of the

10 CFR [Part] 50, Appendix J Type A test (i.e., Containment

Integrated Leakage Rate Test (CILRT)) from 42 months to 66 months.

This test is performed to determine that the total leakage from

containment does not exceed the maximum allowable primary

containment leakage rate (i.e., designated La) at a calculated peak

containment internal pressure (Pa), as defined in 10 CFR [Part] 50,

Appendix J. The primary containment limits the leakage of

radioactive material during and following design bases accidents in

order to comply with the offsite dose limits specified in 10 CFR

[Part] 100. Accordingly, the primary containment is not an accident

initiator, it is an accident mitigator. No physical or operational

changes to the containment structure, plant systems, or components

would be made as a result of the proposed change. Therefore, the

probability of occurrence of an accident previously evaluated is not

increased.

The failure effects that are potentially created by the proposed

one-time TS change have been considered. The relevant components

important to safety which are potentially affected are the

containment structure, plant systems, and containment penetrations.

There are no physical or operational changes to any plant equipment

associated with the proposed TS change. Therefore, the probability

or consequences of a malfunction of equipment important to safety is

not increased.

The proposed change introduces the possibility that primary

containment leakage in excess of the allowable value (i.e., La)

would remain undetected during the proposed 24 month extension of

the interval between the second and third Type A test. The types of

mechanisms which could cause degradation of the primary containment

can be categorized into two types. These are: 1) degradation due to

work which is performed as part of a modification or maintenance

activity on a component or system (i.e., activity-based), or; 2)

degradation resulting from a time-based failure mechanism.

A review of activity-based failure mechanisms has determined

that the potential from degradation due to activity based mechanisms

is minimal.

Regarding the potential for primary containment degradation due

to a time-based mechanism, we have concluded that the PBAPS Local

Leak Rate Test (LLRT) program would identify most types of

penetration leakage. The LLRT program involves measurement of

leakage from Type B and Type C primary containment penetrations as

defined in 10 CFR [Part] 50, Appendix J.

The 10 CFR [Part] 50, Appendix J, Type B tests are intended to

detect local leaks and to measure leakage across pressure containing

or leakage-limiting boundaries other than valves, such as

containment penetrations incorporating resilient seals, gaskets,

expansion bellows, flexible seal assemblies, door operating

mechanism penetrations that are part of the containment system,

doors, and hatches. 10 CFR [Part] 50, Appendix J, Type C testing is

intended to measure reactor system primary containment isolation

valve leakage rates. The frequency of the Type B and Type C testing

is not being altered by the proposed TS change. [However, in an

April 18, 1994 letter, the licensee has requested a 60-day extension

of the Type B and Type C testing.] The acceptance criterion for Type

B and Type C leakage is 0.6 La (i.e., 0.3 % wt/day) which, when

compared to the Type A test acceptance criterion of 0.75 La (i.e.,

0.375 % wt/day), is a significant portion of the Type A test

allowable leakage.

The proposed TS change only extends the interval between two

consecutive Type A tests. The Type B and Type C tests will be

performed as required. The Type B and Type C tests will continue to

be used to confirm that the containment isolation valves and

penetrations have not degraded. Containment system components that

would not be tested are the containment structure itself and small

diameter instrumentation lines. Time-based degradation of any of the

instrumentation lines would most likely be identified by faulty

instrument indication or during instrument calibrations that will be

performed during the PBAPS, Unit 2 refueling outage 10. In examining

the potential for a time-based failure mechanism that could cause

significant degradation of the containment structure, we concluded

that the risk, if any, of such a mechanism is small since the design

requirements and fabrication specifications established for the

containment structure are in themselves adequate to ensure

containment leak tight integrity.

Based on the above evaluation, we have concluded that the

proposed TS change will have a negligible impact on the consequences

of any accident previously evaluated. To support this conclusion, a

review of the PBAPS, Unit 2 CILRT history was performed. This review

identified that the only failure mechanism that has been detected

during the past CILRTs is an activity based component failure, and

that there is no indication of any time-based degradation that would

not be identified during performance of Type B and Type C tests.

Although this review concluded that the risk of undetected

primary containment degradation is not increased, the Individual

Plant Examination (IPE) for PBAPS, Units 2 and 3, was also reviewed

in order to assess the impact of exceeding the primary containment

allowable leakage rate, if a non-mechanistic activity type (i.e.,

time-based) failure were to occur. The IPE included an evaluation of

the effect of various containment leakage sizes under different

scenarios. The IPE results showed that a containment leakage rate of

35% wt/day would represent less than a 5% increase in risk to the

public being exposed to radiation. This evaluation was based on a

study performed by Oak Ridge National Laboratory for light water

reactors that evaluated the impact of leakage rates on public risk.

As stated earlier, the current value of La for PBAPS, Unit 2, is

0.5% wt/day, which is significantly less than the 35% wt/day

discussed in the IPE evaluation.

Therefore, the proposed TS change involving a one-time extension

of the Type A test interval and performing the third Type A test

after the second Appendix J 10-year service period will not involve

an increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change is an increase of a surveillance test

interval and does not make any physical or operational changes to

existing plant systems or components. Primary containment acts as an

accident mitigator not initiator. Therefore, the possibility of a

different type of accident than any previously evaluated or the

possibility of a different type of equipment malfunction is not

introduced.

Therefore, the proposed TS change does not create the

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

previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The total primary containment leakage rate ensures that the

total containment leakage volume will not exceed the value assumed

in the safety analyses at the peak accident pressure. As an added

conservatism, the measured overall leakage rate is further limited

to less than or equal to 0.75 La during performance of periodic

tests to account for possible degradation of the containment leakage

barriers between leakage tests. There is the potential that

containment degradation could remain undetected during the proposed

24 month surveillance interval extension and result in the

containment leakage exceeding the allowable value assumed in safety

analysis. A review of activity-based failure mechanisms has

determined that the potential from degradation due to activity based

mechanisms is minimal.

Regarding the potential for primary containment degradation due

to a time-based mechanism, we have concluded that the PBAPS Local

Leak Rate Test (LLRT) program would identify most types of

penetration leakage. The LLRT program involves measurement of

leakage from Type B and Type C primary containment penetrations as

defined in 10 CFR [Part] 50, Appendix J.

The 10 CFR [Part] 50, Appendix J, Type B tests are intended to

detect local leaks and to measure leakage across pressure containing

or leakage-limiting boundaries other than valves, such as

containment penetrations incorporating resilient seals, gaskets,

expansion bellows, flexible seal assemblies, door operating

mechanism penetrations that are part of the containment system,

doors, and hatches. 10 CFR [Part] 50, Appendix J, Type C testing is

intended to measure reactor system primary containment isolation

valve leakage rates. The frequency of the Type B and Type C testing

is not being altered by the proposed TS change.

Finally, a review of the results of previous PBAPS, Unit 2 CILRT

results concluded that the only failure mechanism which has been

detected during the past CILRTs is activity-based and that there is

no indication of time-based failures that would not be identified

during performance of Type B and Type C tests. Therefore, we have

concluded that the proposed extended test interval would not result

in a non-detectable PBAPS, Unit 2 primary containment leakage rate

in excess of the allowable value (i.e., 0.5% wt/day) established by

the TS and 10 CFR [Part] 50, Appendix J.

Therefore, the proposed TS 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: Government Publications

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

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

Pennsylvania 17105.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller

South Carolina Electric & Gas Company, South Carolina Public

ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: March 11, 1994

Description of amendment request: The proposed amendment would

reduce the allowed outage time for the residual heat removal (RHR)

suction relief valves (SRVs) in accordance with the guidance of Generic

Letter (GL) 90-06.

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 probability or consequences of an accident previously

evaluated is not significantly increased.

This change decreases the allowed outage time of a Low

Temperature Overpressure Protection (LTOP) system. There is no

hardware, software, or operating methodology change, so there is no

increase in probability or consequences. Since the time allowed for

one train of this equipment to be inoperable is shorter, the

probability of an overpressure event not being mitigated has also

been reduced. The consequences will not change unless the system or

operation of the system changes.

2. [The proposed change will not] [c]reate the possibility of a

new or different kind of accident from any previously analyzed.

As this proposed change will not involve any changes to

hardware, software, or operating practices, it cannot create any

possibility of new or different kinds of accidents from those

previously analyzed. The RHR SRVs are intended to provide protection

against a rupture of a pressure boundary from an over-pressure

condition which has the potential to result in core uncovery. The

original design basis of the plant complies with the requirements of

10 CFR 50 Appendix G and uses the RHR SRVs to meet the fracture

toughness requirements of 10 CFR 50 Appendix G. This change only

increases the availability of this protection and does not create

any new or different kinds of accidents.

3. [The proposed amendment does not] [i]nvolve a significant

reduction in a margin of safety.

SCE&G already has administrative controls in place to minimize

the possibility of an overpressure event occurring as well as to

assure that there are two trains of LTOP equipment operable during

the modes when the potential exists for this event. There are

controls to preclude the inadvertent start-up of a Reactor Coolant

Pump or Charging Pump and controls to ensure that both RHR Suction

Isolation Valves for each train are open and remain open except for

testing and maintenance. This alignment is maintained until the RHR

System is realigned for its ECCS function. These controls are

proceduralized in plant operating procedures.

This change does not involve a significant reduction in a margin

of safety as nothing is changed which affects the margin in a

negative direction. The decrease in AOT actually increases the

margin since the allowed time for one train to be inoperable has

been 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: Fairfield County Library,

Garden and Washington Streets, Winnsboro, South Carolina 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: William H. Bateman

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

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

Date of amendment request: May 16, 1994 (TS 94-03)

Description of amendment request: The proposed change would remove

Table 3.3-2, ``Reactor Trip System Instrumentation Response Times,''

and Table 3.3-5, ``Engineered Safety Features Response Times,'' from

the technical specifications and incorporate the limits into the

Updated Final Safety Analysis Report. In addition, references to these

tables in Specifications 3.3.1.1, 3.3.2.1, and 4.3.1.1.3 (for Unit 1)

and 3.3.1, 3.3.2, and 4.3.1.1.3 (for Unit 2) would be removed. A

footnote would be added to Specification 4.3.1.1.3 indicating that

neutron detectors are exempt from response time testing. These changes

have been proposed in accordance with Generic Letter 93-08. A change to

the Bases would indicate that the response time limits would be

maintained in the Updated Final Safety Analysis 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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change does not alter the response time limit

requirements for the reactor trip or engineered safety feature

actuation systems or surveillance testing and frequency. Placing

these limits in the Updated Final Safety Analysis Report (UFSAR)

will ensure the plant design basis is maintained in accordance with

10 CFR 50.59. Since no actual changes to response time limits or

surveillance requirements are involved, the probability or

consequences of an accident are not increased.

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

from any previously analyzed.

The proposed changes does not affect any plant equipment,

functions, or setpoint by relocating response time limits to the

UFSAR. Therefore, the possibility of a new or different kind of

accident is not created.

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

The proposed change will continue to require SQN to maintain the

plant functions at the required setpoints necessary for the design

basis and to support the accident analysis. The margin of safety is

not reduced because there is no change to plant functions and the 10

CFR 50.59 process will continue to ensure the plant design basis is

appropriately maintained.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: May 18, 1994 (TS 94-05)

Description of amendment request: The proposed change would add a

note to the action statement for Limiting Condition for Operation

3.7.7, ``Control Room Emergency Ventilation System,'' indicating that

the provisions of TS 3.0.3 are not applicable while performing actions

associated with a tornado warning.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The control room emergency ventilation system (CREVS) was

designed to ensure control room habitability during accident

conditions. The design basis of SQN does not include an accident

creating a contaminated air condition concurrent with a tornado. The

ability of the CREVS to perform its design function has not been

affected by this change. The proposed change will not increase the

possibility or consequences of an accident.

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

from any previously analyzed.

An accident involving a contaminated air condition and a tornado

have been analyzed as part of the SQN design basis. Both accidents

are assumed to occur independently. This change does not create a

new or different accident not previously analyzed.

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

The design basis of the CREVS is not impacted by this TS change.

There is no change in any assumptions made in the Final Safety

Analysis Report. Therefore, there is no reduction in the margin of

safety as a result of this change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: March 19, 1994; superseded May 16, 1994

(TS 93-04)

Description of amendment request: The proposed change would clarify

and consolidate the technical specifications (TS) regarding the dual

function of the containment vacuum relief system (i.e., the vacuum

relief and containment isolation functions). The proposed changes would

revise TS 3/4.6.6, ``Vacuum Relief Valves,'' to indicate the actions

that would be required should one or more vacuum relief (VR) lines be

incapable of performing its containment isolation function or incapable

of performing its VR function. In addition, the testing requirements

would be revised to add specific requirements and reflect the inservice

test (IST) program by relocating the testing requirements from TS

4.6.3.2.d and Table 3.6-2 to the new TS 4.6.6 (and to Sequoyah's IST

program). Other proposed changes affect Bases 3/4.6.6 section and TS

index pages to reflect the proposed changes indicated above. This

proposed change was originally noticed on May 12, 1993 (58 FR 28060),

which is superseded by this notice.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

TVA's proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change does not increase the probability of an accident

since the proposed change does not affect any plant systems,

equipment, or components. The dual design functions of SQN's

containment vacuum relief (VR) system (i.e., provide containment VR

and containment isolation) are not affected. The consequences of an

event are not significantly increased by the proposed increase in

allowed outage time from 4 hours to 72 hours for returning an

inoperable VR system to operable status. The probability of an event

during the relatively short duration of the TS completion times, in

conjunction with the redundancy provided in the design of the

system, provide sufficient assurance that the VR lines are available

for mitigating an accident or abnormal event.

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

from any previously analyzed.

No physical modification is being made to any plant hardware or

plant operating setpoints, limits, or operating procedures as a

result of this change. TVA's proposed change provides a TS

improvement that clarifies the TS requirements associated with the

dual design function of SQN's VR system. The proposed change removes

the potential for creating a conflict between Specification 3/4.6.3,

``Containment Isolation Valves,'' and Specification 3/4.6.6,

``Vacuum Relief Valves.''

The proposed change does not alter any accident analysis or any

assumptions used to support the accident analyses. The containment

leakage assumptions used to determine offsite dose limits for

compliance with 10 CFR 100 are not affected. The analysis that

supports the containment VR system also remains unchanged. The

proposed 72-hour and 1-hour completion times for returning an

inoperable VR line to operable status are consistent with the NUREG-

1431 and NUREG-1433. Consequently, the proposed change does not

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

any previously analyzed.

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

The margin of safety provided by the design of SQN's containment

VR system remains unchanged. TVA's proposed change does not affect

the VR function or the containment isolation function that currently

exists in SQN TSs. The proposed change eliminates the potential for

conflicting requirements within SQN TSs and ensures that the proper

action is taken to preserve these dual design functions while the

plant is in Modes 1, 2, 3, or 4. TVA's proposed change provides a TS

improvement that combines these functional requirements into a

single specification. Both VR and containment isolation requirements

will continue to be provided. Accordingly, the proposed change does

not involve a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

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: February 14, 1993

Brief description of amendments: The proposed amendments would

revise the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2,

technical specifications (TS) by (1) changing the allowable value for

Unit 2 overtemperature N-16 and pressurizer pressure-low setpoints, (2)

deleting Equation 2.2-1 from TS 2.2.1, and (3) administrative changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of a previously evaluated

accident.

Overtemperature N-16, Unit 2

Incorporation of the increased temperature uncertainties

reported by Rosemount will change the Allowable Value of the

Overtemperature N-16 trip function. The change does not affect the

Safety Analysis Limits assumed in the accident analysis. Because the

change only impacts the Allowable Value for a setpoint and does not

affect any system designs or operations, the change does not

increase the probability of an accident. Although the Allowable

Value is changed in the conservative direction, the change assures

that, considering the newly identified transmitter uncertainty, the

trip actuates prior to the conditions assumed in the accident

analyses. As such, there is no impact on the consequences of any

accidents previously evaluated.

Pressurizer Pressure - Low, Unit 2

The added uncertainties change the Allowable Value of the Unit 2

Pressurizer Pressure-Low Reactor Trip function. The change does not

affect the Safety Analysis Limits assumed in the accident analysis.

Because the change only impacts the Allowable Value for a setpoint

and does not affect the system design or operations, the change does

not increase the probability of an accident. Although the Allowable

Value is changed in the conservative direction, the change assures

that, considering the newly identified transmitter uncertainty, the

trip actuates prior to the conditions assumed for the accident

analyses. As such, there is no impact on the consequences of any

accidents previously evaluated.

Equation 2.2-1

The changes to Specifications 2.2.1 and 3.3.2, to Tables 2.2-1

and 3.3-3, and to the bases sections will require recalibration of

the channel and removal of any accumulated errors in any function

whose ``as found'' setpoint is found to be less conservative than

its allowable value. These changes delete a potentially less

conservative option and will result in actual channel operation

closer to the nominal setpoint and within the allowable value band.

These changes will in effect validate one of the assumptions made in

the accident analysis and will not increase the probability or

consequences of any accident evaluated in the Safety Analysis

Report.

Administrative Changes

The changes to combine the Unit 1 and Unit 2 line items into a

dual Unit line if the Trip Setpoint and Allowable Value values are

the same is administrative and meant as a human factors improvement

for operator convenience. The change does not affect the operation

of any equipment, the operating point of any equipment, nor any

equipment hardware and thus does not increase the probability or

consequences of any accident evaluated in the Safety Analysis

Report.

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

or different kind of accident from any previously analyzed accident.

Overtemperature N-16 and Pressurizer Pressure - Low, Unit 2

As the proposed amendment changes only the Unit 2 Allowable

Values of the Overtemperature N-16 reactor trip and the Pressurizer

Pressure-Low reactor trip and does not have any physical effect on

the transmitter or circuitry, there are no new or different types of

accident introduced.

Equation 2.2-1

Deletion of this equation and its associated action statements,

definitions and values does not introduce any physical changes to

any systems, structures, or components. The change merely assures

that setpoints which are less conservative than their Allowable

Value are recalibrated prior to being declared operable. These

changes do not introduce any new credible failure modes which may

create the possibility of a new or different accident.

Administrative Changes

Combining line items for Unit 1 and Unit 2 into a dual Unit

entry for administrative purposes does not introduce any new

credible failure modes which may create the possibility of a new or

different accident.

3. The proposed change does not involve a significant reduction

in the margin of safety.

Overtemperature N-16 and Pressurizer Pressure - Low, Unit 2

Incorporation of the added temperature uncertainties of the

Rosemount transmitters assures that the safety analysis limits

assumed in the accident analyses for Overtemperature N-16 and

Pressurizer Pressure-Low reactor trip functions for Unit 2 are met.

There is no change in the acceptance criteria or the results of

these analyses due to this change. Thus there is no effect on the

margin of safety.

Equation 2.2-1

Deletion of Equation 2.2-1, related actions and associated

definitions and values, merely eliminates one option to assure that

the safety analysis assumptions are met. This option is not

presently in use and the accident analyses assumptions have been and

will continue to be met using the other option (to re-calibrate

channels prior to restoring operability). Thus the margin of safety

is unaffected.

Administrative Changes

Combining the Unit 1 and Unit 2 line items of Table 2.2-1 for

RTS [Reactor Trip Systems] functions and of Table 3.3-3 for ESFAS

[Engineered Safety Features Actuation System] functions into dual

unit entries does not change the Trip Setpoint or the Allowable

Value for the functions. The margin of safety is unaffected.

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, 701 South Cooper, 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

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: April 19, 1994

Description of amendment request: The proposed amendment revises

Technical Specification 6.2.2.g to reflect a title designation change

within the Wolf Creek Nuclear Operating Corporation (WCNOC)

organization. The title of Supervisor Operations is being changed to

Assistant Manager Operations. The title change does not represent any

change in reporting relationships, job responsibilities, or overall

organizational commitments.

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 proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

This change involves an administrative change to the WCNOC

organization and to the position title and as such has no effect on

plant equipment or the technical qualification of plant personnel.

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

or different kind of accident from any previously evaluated.

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

different kind of accident from any previously evaluated. This

change is administrative in nature and does not involve any change

to installed plant systems or the overall operating philosophy of

Wolf Creek Generating Station.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change does not involve a significant reduction in

a margin of safety. This change does not involve any changes in

overall organizational commitments. A position title change alone

does not reduce any margin of safety.

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

this review, it appears that the three standards of 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 locations: Emporia State University,

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

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

66621

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: Theodore R. Quay

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: April 19, 1994

Description of amendment request: The proposed amendment revises

Technical Specification Table 3.6-1, ``Containment Isolation Valves,''

by deleting reference to two (2) valves. The Technical Specification

change reflects a planned modification which removes the essential

service water (ESW) containment air cooler return line isolation valve

bypass valves and associated piping.

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.

After the design modification is completed the ESW Containment

Penetrations will be provided with stainless steel isolation valves,

which will be provided with automatic SIS [safety injection signal]

actuation signals to open automatically to provide required cooling

water flow to the Containment Air Coolers following a LOCA [loss-of-

coolant accident] or MSLB [main steamline break]. Replacement of the

current carbon steel isolation valves with stainless steel valves

and removing the unnecessary bypass lines and bypass isolation

valves will reduce the amount of seat leakage currently experienced

with these valves.

The probability of occurrence of a previously evaluated accident

is not increased because this modification does not introduce any

new potential accident initiating conditions. The consequences of an

accident previously evaluated is not increased because the ability

of containment to restrict the release of any fission product

radioactivity to the environment will not be degraded by this

modification.

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

or different kind of accident from any previously evaluated.

The proposed modification will reduce the number of containment

isolation valves and replace several carbon steel isolation valves

with stainless steel valves, which will be less susceptible to

erosion and corrosion. Thus, potential system leakage will be

reduced by this modification, while valve reliability will be

enhanced. The new valves are designed to the original ESW System

requirements, and removal of the bypass lines and bypass isolation

valves will not result in a malfunction of any other plant

equipment. Therefore, this proposed modification will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The removal of the bypass lines and bypass isolation valves will

not adversely affect containment isolation capability for credible

accident scenarios. Due to a previous design change, the bypass

lines are no longer required to ensure adequate cooling flow to the

Containment Air Coolers. In addition, the operability and

reliability of the remaining isolation valves will be enhanced by

replacing the current carbon steel valves with stainless steel

valves.

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 locations: Emporia State University,

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

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

66621

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: Theodore R. Quay

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

Carolina Power & 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: May 11, 1994

Brief description of amendment request: The amendment would allow

reduced power operation as a function of reactor coolant system (RCS)

total flow rate for flow rate reductions of up to 5 percent below the

currently specified flow rate. Operation will be allowed at total flow

rates slightly lower than (293,540 gpm X (1.0 plus C1)) if rated

thermal power (RTP) is reduced by 1.5 percent for each one percent that

RCS total flow is less than this rate. This change would provide for

needed operational margin and flexibility without the unnecessary

penalty of a large power reduction.

Date of publication of individual notice in Federal Register: May

25, 1994 (59 FR 27079)

Expiration date of individual notice: June 24, 1994

Local Public Document Room location: Cameron Village Regional

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

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

NuclearPower Station, Unit 1, New London County, Connecticut

Date of amendment request: May 27, 1994

Description of amendment request: The amendment would add a new

section to Technical Specification Section 6.17 and would require that

procedures be in place to provide for monitoring and sampling of

emergency service water (ESW) discharge flow during accident conditions

when a positive differential pressure cannot be maintained between ESW

and low pressure coolant injection (LPCI) in the LPCI heat exchangers.

Date of publication of individual notice in Federal Register: June

7, 1994 (59 FR 29448)

Expiration date of individual notice: July 7, 1994

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

Notice Of Issuance Of Amendments To Facility Operating LIcenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

Street, NW., Washington, DC 20555, and at the local public document

rooms for the particular facilities involved.

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

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

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

Date of application for amendments: February 18, 1994, as

supplemented by letter dated May 16, 1994

Brief description of amendments: These amendments modify Technical

Specification (TS) Figure 3.2-1, ``REACTOR COOLANT COLD LEG vs CORE

POWER LEVEL,'' of TS 3/4.2.6, ``REACTOR COOLANT COLD LEG TEMPERATURE,''

for Units 1 and 3 to include the cold leg temperature between 552 deg.F

and 562 deg.F at core power levels between 90 percent and 100 percent

within the AREA OF ACCEPTABLE OPERATION. Also, the proposed amendments

modify TS 3/4.1.1.4, ``MINIMUM TEMPERATURE FOR CRITICALITY,'' and BASES

3/4.1.1.4, ``MINIMUM TEMPERATURE FOR CRITICALITY,'' to allow the

minimum temperature for criticality to be established at 545 deg.F,

rather than the current value of 552 deg.F, to establish the

surveillance temperature at 552 deg.F, rather than the current

557 deg.F, and to clarify the BASES for this TS.

Date of issuance: June 7, 1994

Effective date: NPF-41 and NPF-51, prior to startup from the next

refueling outage; NPF-74, no later than 45 days from the date of

issuance.

Amendment Nos.: 77, 63, and 49

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

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 30, 1994 (59 FR

14886) The additional information contained in the May 16, 1994, letter

was clarifying in nature, was within the scope of the initial notice,

and did not affect the NRC staff's proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated June 7, 1994.No

significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County,North Carolina

Date of application for amendments: April 14, 1994, as supplemented

on May 16, 1994.

Brief description of amendments: The amendments change the

Technical Specifications (TS) to relocate the Instrument Response Time

Tables to the Updated Final Safety Analysis Report in accordance with

NRC Generic Letter 93-08.

Date of issuance: May 31, 1994

Effective date: May 31, 1994

Amendment Nos.: 171 and 202

Facility Operating License Nos. DPR-71 and DPR-62. Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: April 26, 1994 (59 FR

21785) The May 16, 1994, letter provided clarifying information that

did not change the initial no significant hazards consideration

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

contained in a Safety Evaluation dated May 31, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

Date of application for amendments: August 20, 1993, as

supplemented by letters dated December 27, 1993, March 22, 1994, and

May 31, 1994.

Brief description of amendments: The amendments delete Technical

Specification Section 3/4.6.1.5, ``Primary Containment Structural

Integrity'' which includes Surveillance Requirements for the Primary

Containment Tendons and adds a Technical Specification requirement to

establish, implement, and maintain a comprehensive containment tendon

program. The containment tendon program is based on Regulatory Guide

1.35, Rev. 3, and is titled ``Inservice Inspection Program for Post

Tensioning Tendons.'' The new program will allow the Unit 1 and 2

containments to be tested as twin containments.

Date of issuance: June 3, 1994

Effective date: June 3, 1994

Amendment Nos.: 100 and 84

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59746) The supplemental information submitted December 27, 1993,

March 22, 1994, and May 31, 1994, contained clarifying information

related to the original request, and did not change the no significant

hazards finding. The Commission's related evaluation of the amendments

is contained in a Safety Evaluation dated June 3, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Public Library of Illinois

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

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, Rock Island County,

Illinois

Date of application for amendments: December 20, 1993

Brief description of amendments: The amendments increase the

minimum critical power ratio (MCPR) from 1.06 to 1.07 for Quad Cities,

Units 1 and 2, as a result of the planned implementation of GE 8x8NB-3

fuel for Cycle 14 of each unit.

Date of issuance: June 10, 1994

Effective date: June 10, 1994

Amendment Nos.: 146 and 142

Facility Operating License Nos. DPR-29 and DPR-30. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 17, 1994 (59

FR 10003) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 10, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of application for amendment: February 25, 1994

Brief description of amendment: The amendment adds a new Technical

Specification 3/4.7.12, ``Ultimate Heat Sink'' and its associated Bases

Section 3/4.7.12.

Date of Issuance: May 31, 1994

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

30 days.

Amendment No.: 172

Facility Operating License No. DPR-61. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (59 FR

17596) The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated May 31, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: December 6, 1993

Brief description of amendment: The amendment revises the Technical

Specifications (TSs) to provide several temporary one-time changes that

are necessary to support the fuel out, chemical decontamination program

that is currently scheduled for the upcoming 1995 refueling outage.

Specifically, the amendment revises the definition of the cold shutdown

condition in TS 1.2.1 by changing the upper limit of Tavg for the

cold shutdown condition from 200 deg.F to 250 deg.F. The amendment also

revises the definition of the hot shutdown condition in TS 1.2.2 by

changing the lower limit of Tavg for the hot shutdown condition

from greater than 200 deg.F to greater than 250 deg.F.

Date of issuance: June 9, 1994

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

30 days.

Amendment No.: 170

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 16, 1994 (59

FR 7687) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 9, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

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

Date of application for amendments: January 27, 1994

Brief description of amendments: The amendments would eliminate the

humidity control functions of the containment purge (VP) system

humidistats by deleting the surveillance requirement (SR) for periodic

verification of automatic isolation of the VP system on a high relative

humidity (RH) test signal and heater failure from the existing SR for

Catawba Units 1 and 2.

Date of issuance: May 25, 1994

Effective date: May 25, 1994

Amendment Nos.: 118 and 112

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

revised the Technical Specifications.

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

10005) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 25, 1994.No significant

hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: April 29, 1993, as supplemented

May 16, 1994

Brief description of amendments: The amendments delete License

Condition 2.C.(20) from Facility Operating License NPF-35 for Unit 1,

and License Condition 2.C.(11) from Facility Operating License NPF-52

for Unit 2. These conditions address engine teardown and inspection

required following the crankshaft failure of an Enterprise emergency

diesel generator at the Shoreham Nuclear Plant.

Date of issuance: June 2, 1994

Effective date: June 2, 1994

Amendment Nos.: 119/113

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 26, 1993 (58 FR

30192) The May 16, 1994, letter provided additional information that

did not change the scope of the April 29, 1993, application and

proposed initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 2, 1994.No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: January 13, 1993, as

supplemented January 28, February 17, and April 26, 1993.

Brief description of amendments: The amendments revise Technical

Specification Table 2.2.1, Sections 3/4.1.2.5, 3/4.1.2.6, 3/4.5.1.1, 3/

4.5.5, and their associated Bases, and Technical Specification 6.9.1.9,

to relocate the values of certain cycle-dependent limits from the

Technical Specifications to the Core Operating Limits Report.

Date of issuance: May 31, 1994

Effective date: May 31, 1994

Amendment Nos.: 143 and 125

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 4, 1993 (58 FR

41503) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 31, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: March 23, 1994, as supplemented

April 14, May 11, and May 17 (two letters) 1994.

Brief description of amendments: The amendments relating to the

March 23, 1994, application revise Technical Specification (TS) 6.9.2,

``Core Operating Limits Report,'' (COLR) to include a reference to a

Duke Power Company Topical Report describing an analytical method for

determining the core operating limits. Specifically, the amendments

add: ``(4) DPC-NE-1004A, Nuclear Design Methodology Using CASMO-3/

SIMULATE-3P,'' to TS 6.9.2.

The May 11, 1994, letter added a statement to TS 6.9.2 that the

approved methods used to determine the core operating limits given in

TS 6.9.1 are specified in the COLR. The May 11 and 17, 1994, letters

provided information regarding Duke Power's transition from the EPRI-

NODE-P based methodology to the simulate methodology. Revision 1 to the

COLR for Oconee 1 Cycle 16 was submitted by letter dated May 17, 1994.

The April 14, 1994, letter revised the TS Table of Contents to

delete reference to Table 4.4-1. This table was removed from the TS by

an amendment issued on September 16, 1993.

Date of Issuance: June 8, 1994

Effective date: June 8, 1994

Amendment Nos.: 206, 206, and 203

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 28, 1994 (59 FR

22007) The April 14, May 11, and May 17 (two letters), 1994, letters

provided additional information that did not change the scope of the

March 23, 1994, application and the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

June 8, 1994. No significant hazards consideration comments received:

No

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 23, 1993

Brief description of amendment: The amendment revised the Technical

Specifications in accordance with Generic Letter 93-05, ``Line Item

Technical Specification Improvements To Reduce Surveillance

Requirements For Testing During Power Operation'' for radiation

monitors, pressurizer heaters, reactor coolant isolation valves, and

auxiliary feedwater pumps.

Date of issuance: June 6, 1994

Effective date: June 6, 1994

Amendment No.: 96

Facility Operating License No. NPF-38. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 16, 1994 (59

FR 7689) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 6, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Burke County, Georgia

Date of application for amendments: December 30, 1993

Brief description of amendments: The proposed change would allow a

one time extension of the allowable outage time for each residual heat

removal (RHR) pump from 3 to 7 days to allow modifications to the RHR

system while the plant is in Mode 1.

Date of issuance: May 31, 1994

Effective date: May 31, 1994

Amendment Nos.: 72 and 51

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

revised the Technical Specifications.

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

10007) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 31, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Burke County, Georgia

Date of application for amendments: November 19, 1993

Brief description of amendments: The amendments modify Technical

Specification Table 3.3-2, Engineered Safety Features Actuation System

Instrumentation, modifying the Mode for which Item 6.e, ``Trip of All

Main Feedwater Pumps, Start Motor-Driven Pumps,'' is required to be

operable.

Date of issuance: June 1, 1994

Effective date: June 1, 1994

Amendment Nos.: 73 and 52

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58

FR 67847) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June, 1, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Burke County, Georgia

Date of application for amendments: March 1, 1994

Brief description of amendments: The amendments modify Technical

Specification (TS) 3.2.4, ``Quadrant Power Tilt Ratio,'' by adding an

exception to the requirements of TS 3.0.4.

Date of issuance: June 1, 1994

Effective date: June 1, 1994

Amendment Nos.: 74 and 53

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (59 FR

17599) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 1, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

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

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

Date of application for amendment: February 7, 1994

Brief description of amendment: The amendment revises the plant

Technical Specifications (TS) to require the Three Mile Island, Unit 1

(TMI-1) annual radioactive effluent release report for the previous

calendar year be submitted by May 1 of each year. The current TS

requires the TMI-1 report be submitted within 60 days after January 1

of each year. Changing the TMI-1 due date to May 1 enables the licensee

to combine the reports for TMI-1 and TMI-2 into a single report with a

common due date.

Date of Issuance: June 10, 1994

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

30 days.

Amendment No.: 185

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (59 FR

17600) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 10, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

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

Matagorda County, Texas

Date of amendment request: May 27, 1993, as supplemented by letter

dated April 18, 1994.

Brief description of amendments: The amendments upgrade the fuel

used in the South Texas Project reactors to Westinghouse VANTAGE 5

Hybrid (V5H) design and implement several analytical and operational

upgrades into the South Texas Project Updated Final Safety Analysis

Report. The amendments modify related setpoints, limiting conditions

for operation, surveillance requirements, design features information,

and associated bases in the following specifications: TS Table 2.2-1,

``Reactor Trip System Instrumentation Trip Setpoints,'' TS Figure 3.1-

1, ``Required Shutdown Margin for Modes 1 and 2,'' TS Figure 3.1-2,

``Required Shutdown Margin for Mode 5,'' TS Figure 3.1-2a, ``MTC versus

Power Level,'' TS 3/4.2.5, ``Power Distribution Limits - DNB

Parameter,'' TS Table 3.3-4, ``Engineered Safety Features Actuation

System Instrumentation Trip Setpoints,'' TS 3/4.6.1.1, ``Primary

Containment - Containment Integrity,'' TS 3/4.6.1.2, ``Containment

Systems - Containment Leakage,'' TS 3/4.6.1.3, ``Containment Systems -

Containment Air Locks,'' TS 3/4.6.1.5, ``Containment Systems - Air

Temperature,'' TS 3/4.7.1.2, ``Plant Systems - Auxiliary Feedwater

System,'' TS 5.2.1, ``Containment - Configuration,'' TS 5.3.1,

``Reactor Core - Fuel Assemblies,'' TS 5.6.1, ``Fuel Storage -

Criticality,'' and adds TS Figure 5.6-7, ``Minimum IFBA Content for In-

Containment Rack Fuel Storage.''

Date of issuance: May 27, 1994

Effective date: May 27, 1994, to be implemented prior to completion

of Unit 1 REO5

Amendment Nos.: Unit 1 - Amendment No. 61; Unit 2 - Amendment No.

50

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 7, 1993 (58 FR

36436) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 27, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Wharton County Junior College,

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

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,

Nuclear Plant, Unit No. 2, Berrien County, Michigan

Date of application for amendment: March 9, 1994, as supplemented

April 13, 1994.

Brief description of amendment: The amendment revises the Technical

Specifications to allow a one-time extension for Type B and C leak rate

tests. The Commission had previously granted a one-time schedular

exemption from the requirements in 10 CFR Part 50, Appendix J,

paragraphs III.D.2.(a) and III.D.3. The exemption extends the maximum

allowable time between tests by 150 days.

Date of issuance: June 1, 1994

Effective date: June 1, 1994

Amendment No.: 162

Facility Operating License No. DPR-74. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: April 28, 1994 (59 FR

22009). The April 13, 1994, supplemental letter provided clarifying

information that was within the scope of the April 28, 1994, notice.

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

Safety Evaluation dated June 1, 1994.No significant hazards

consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

Date of application for amendment: January 14, 1992 (TS 300)

Brief description of amendments: The amendments add requirements to

the Browns Ferry Units 1 and 3 Technical Specifications to ensure

thermal-hydraulic stability, consistent with guidance provided by NRC

Bulletin 88-07 ``Power Oscillations in Boiling Water Reactors,'' and

Supplement 1 to that Bulletin.

Date of issuance: May 31, 1994

Effective date: May 31, 1994

Amendment Nos.: 206 and 179

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

Date of initial notice in Federal Register: April 15, 1992 (57 FR

13138) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 31, 1994.No significant

hazards consideration comments received: None

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

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: February 8, 1994 (TS 93-14)

Brief description of amendments: The amendments increase the

pressure setpoint for the motor driven auxiliary feedwater pumps

switchover from the condensate storage tank to the essential raw

cooling water supply.

Date of issuance: May 27,1994

Effective date: May 27, 1994

Amendment Nos.: 183 and 175

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

revise the technical specifications.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12368) The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated May 27, 1994.No significant

hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: July 14, 1993

Brief description of amendment: This amendment revises Sections 3.6

and 4.6 of the Technical Specifications to incorporate reactor coolant

system leakage detection requirements to address Generic Letter 88-01

``NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in

BWR Austenitic Stainless Steel Piping.''

Date of issuance: June 1, 1994

Effective date: June 1, 1994

Amendment No.: 139

Facility Operating License No. DPR-28. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12370) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 1, 1994 No significant

hazards consideration comments received: No

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301.

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

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

and regulations. The Commission has made appropriate findings as

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

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

Because of exigent or emergency circumstances associated with the

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

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

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,

and at the local public document room for the particular facility

involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By July 22, 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 intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

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

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

effect.

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

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

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

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

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

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

Commonwealth Edison Company, Docket No. STN 50-456, Braidwood

Station, Unit No. 1, Will County, Illinois

Date of application for amendments: April 25, 1994, as supplemented

April 28, 1994, April 30, 1994, May 2, 1994, May 4, 1994, and May 6,

1994.

Brief description of amendments: The amendment revises Braidwood,

Unit 1, technical specifications (TSs) in Appendix A to the operating

license by adding additional surveillance and operating requirements to

Section 4.4.5.2, ``Steam Generator Tube Sample Selection and

Inspection; Section 4.4.5.4, ``Acceptance

Criteria; Section 4.4.5.5, ``Reports; and Section

3.4.6.2. This amendment is applicable only for 100 calendar days from

the date of issuance, not counting any time when the Thot

temperature is below 500 deg.F. These changes revise the existing steam

generator tube repair criteria to allow usage of the voltage-based

criteria identified by the staff in draft NUREG-1477 as the interim

plugging criteria (IPC). Additionally, a footnote is added to TS 3.4.8

to limit the dose equivalent iodine-131 concentration to 0.35

microcuries per gram of coolant for the limited time period cited

above. The Unit 1 Bases are revised to be consistent with the changes

cited above.

Date of issuance: May 7, 1994

Effective date: May 7, 1994

Amendment No.: 50

Facility Operating License No. NPF-72. The amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes. The NRC published a public

notice of the proposed amendment, issued a proposed finding of no

significant hazards consideration and requested that any comments on

the proposed no significant hazards consideration be provided to the

staff by the close of business on May 5, 1994. The notice was published

in the Herald News and the Morris Daily Herald on May 3, 1994. The

Commission's related evaluation of the amendment, finding of emergency

circumstances, and final determination of no significant hazards

consideration are contained in a Safety Evaluation dated May 7, 1994.

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

Austin, One First National Plaza, Chicago, Illinois 60690

Local Public Document Room location: Wilmington Township Public

Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

NRC Project Director: James E. Dyer

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

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

Date of application for amendment: May 19, 1994

Brief description of amendment: The amendment revises the

surveillance requirements in TS 3.3.9.3 and 3.3.10.3, to change the

neutron power limits i.e., 105 neutron counts per second (cps) and

1E-6 amperes (amps) indications on the source and intermediate range

instruments, respectively, for verifying overlap between them.

Date of issuance: May 27, 1994

Effective date: May 27, 1994

Amendment No.: 150

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

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: No. The Commission's related

evaluation of the amendment and the final determination of no

significant hazards consideration comments are contained in a Safety

Evaluation dated May 27, 1994.

Attorney for the Licensee: Harold F. Reis, Esquire, Newman and

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

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

NRC Project Director: Herbert N. Berkow

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

MillstoneNuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of application for amendment: May 27, 1994, as supplemented

June 1, 1994.

Brief description of amendment: The amendment revises the Technical

Specifications (TS) by adding a footnote to Tables 3.3-3, 3.3-4 and

3.3-5 of the Millstone Unit No. 2 TS denoting that the operability of

the automatic initiation logic for the auxiliary feedwater system will

rely on operator action for the remainder of Cycle 12.

Date of issuance: June 7, 1994

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

within 30 days.

Amendment No.: 176

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

Technical Specifications.Public comments requested as to proposed no

significant hazards consideration: No. The Commission's related

evaluation of the amendment, finding of emergency circumstances, and

final determination of no significant hazards consideration are

contained in a Safety Evaluation dated June 7, 1994.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz

Dated at Rockville, Maryland, this 15th day of June 1994.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/IIOffice of Nuclear Reactor

Regulation

[Doc. 94-15025 Filed 6-21-94 8:45 am]

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