Biweekly Notice

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 August 8, 1994, through August 19, 1994. The

last biweekly notice was published on August 17, 1994 (59 FR 42332).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

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

intervene is discussed below.

By September 30, 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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: June 9, 1994

Description of amendment request: The proposed amendment would

increase the allowed out-of-service time to increase from 7 days to 14

days for the automatic depressurization system (ADS), the high pressure

coolant injection (HPCI) system and the reactor core isolation cooling

(RCIC)system. The proposed change includes a change to Section 4.5.H,

``Maintenance of Filled Discharge Pipe'' to reflect Amendment 149,

issued September 28, 1993.

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 Operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Safety criteria used to determine the acceptability of extending

continued operation with one ADS valve, the HPCI or RCIC system out-

of-service (OOS) is consistent with Pilgrim's licensing basis. For

example, events with the expected frequency of occurrence greater

than once-per-reactor lifetime are required to meet the transient

MCPR [minimum critical power ratio] thermal limit: more than 99.9%

of the fuel rods are expected to avoid boiling transition. Very low

probability events, such as a LOCA [loss-of-coolant accident], are

required to satisfy the criteria of 10CFR50.46: the primary

criterion being that the Peak Cladding Temperatures (PCT) be

maintained less than 2200 deg.F.

For intermediate frequency events, e.g. safe shutdown in the

event of a fire, 10CFR50 Appendix R involves a ``no fuel damage''

criterion. To evaluate these types of events, the GE [General

Electric] SAFER/GESTR-LOCA licensing methodology was used to

calculate the system responses and PCTs.

Analyses performed by Pilgrim's NSSS [nuclear steam supply

system] vendor, General Electric, [***] for various limiting-case

scenarios involving ADS, HPCI, or RCIC out-of-service situations

demonstrated 10CFR50.46 limits (i.e. a PCT less than 2200 deg.F)

were met. (The most severe PCT was 1500 deg.F). The core damage

frequency analysis for Pilgrim is unchanged by operating Pilgrim in

accordance with this proposed amendment. The 14 day OOS for HPCI,

RCIC and ADS also conforms to the 00S time for these systems found

in BWR [boiling-water reactor] Standard Technical Specifications.

Hence, increasing the allowed 00S time from 7 to 14 days does not

result in a challenge to fuel cladding integrity or BWR Standard

Technical Specifications, and operating Pilgrim in accordance with

the proposed amendment will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The removal of the association between LPCI [low-pressure

coolant injection] and Core Spray system testing and surveilling

their filled discharge pipes is an administrative change because the

specified surveillance frequency is unchanged. This proposed change

reflects Amendment 149, issued by the NRC September 28,

1993, and is proposed to ensure consistency between Pilgrim's

Technical Specification sections. This administrative change will

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. The operation of Pilgrim 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.

As discussed above, a variety of limiting-case scenarios were

analyzed to demonstrate the effects of increasing the 00S time for

one ADS valve, the HPCI system, or the RCIC system. The conclusion

of the analyses is that this proposed change does not violate

Pilgrim's licensing basis or 10CFR50.46 requirements.

Some scenarios result in elevated PCTs, but they are still

significantly below the 10CFR50.46 limit of 2200 deg.F. Therefore,

since the licensing-basis and code required PCT continues to be met

and because the proposed change comports the requirements of

BWR Standard Technical Specifications, operating Pilgrim in

accordance with the proposed amendment will not create the

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

previously evaluated.

As discussed in above question 1, the proposed change to section

4.5.H.1 is administrative and does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

Certain scenarios analyzed for system unavailability result in

evaluated PCTs. However, these elevated PCTs are significantly below

the 10CFR50.46 limit of 2200 deg.F. Therefore, there is no reduction

in the safety margin for PCT resulting from the change from 7 to 14

days. The proposed change also corresponds to the requirements of

BWR Standard Technical Specifications concerning 00S for HPCI, RCIC

and ADS. Therefore, operating Pilgrim Station in accordance with

this proposed amendment does not involve a significant reduction in

a margin of safety.

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

this review, it appears that the three standards of 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler

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 amendment request: July 22, 1994

Description of amendments request: The proposed amendment would

implement a performance based assessment program, including

corresponding organizational and functional changes. Specifically, the

changes affect the Independent Review (IR) function, the independent

assessment of plant activity and the Independent Safety Engineering

Group. These functions will be performed by the proposed Nuclear

Assessment Section (NAS). The NAS would perform internal evaluations

and assessment activities and serve as plant management's staff for the

objective oversight of plant performance relating to nuclear safety,

reliability, and quality. The NAS's fundamental role will be to: (1)

assist plant management in the early identification of issues which may

prevent the plant from achieving quality performance on a sustained

basis; and (2) ensure effective correction of deficiencies.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated because it is a programmatic and administrative

change which does not physically alter any safety-related systems,

nor does it affect the way in which any safety-related systems

perform their functions. Since the design of the facility and system

operating parameters are not changing, the proposed amendment does

not involve an increase in the probability or consequences of any

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. As stated in Item 1, the proposed amendment is a

programmatic and administrative change which does not physically

alter any safety-related systems; nor does it affect the way in

which any safety-related systems perform their functions. Since the

design of the facility and system operating parameters are not

changing, the proposed amendment does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety because it is a programmatic and

administrative change which provides assurance that plant operations

continue to be conducted in a safe manner through the performance

based assessment programs. As stated in Item 1, the proposed

amendment does not physically alter any safety-related systems; nor

does it affect the way in which any safety-related systems perform

their functions. Since the design of the facility and system

operating parameters are not changing, the proposed amendment does

not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 22, 1994

Description of amendment request: The proposed amendment would

implement a performance based assessment program, including

corresponding organizational and functional changes. Specifically, the

changes affect the Independent Review (IR) function, the independent

assessment of plant activity and the Independent Safety Engineering

Group. These functions will be performed by the proposed Nuclear

Assessment Section (NAS). The NAS would perform internal evaluations

and assessment activities and serve as plant management's staff for the

objective oversight of plant performance relating to nuclear safety,

reliability, and quality. The NAS's fundamental role will be to: (1)

assist plant management in the early identification of issues which may

prevent the plant from achieving quality performance on a sustained

basis; and (2) ensure effective correction of deficiencies.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated because it is a programmatic and administrative

change which does not physically alter any safety-related systems,

nor does it affect the way in which any safety-related systems

perform their functions. Since the design of the facility and system

operating parameters are not changing, the proposed amendment does

not involve an increase in the probability or consequences of any

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. As stated in Item 1, the proposed amendment is a

programmatic and administrative change which does not physically

alter any safety-related systems; nor does it affect the way in

which any safety-related systems perform their functions. Since the

design of the facility and system operating parameters are not

changing, the proposed amendment does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety because it is a programmatic and

administrative change which provides assurance that plant operations

continue to be conducted in a safe manner through the performance

based assessment programs. As stated in Item 1, the proposed

amendment does not physically alter any safety-related systems; nor

does it affect the way in which any safety-related systems perform

their functions. Since the design of the facility and system

operating parameters are not changing, the proposed amendment does

not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 28, 1994

Description of amendment request: The proposed amendment would

change TS Sections 5.3.1.3, 5.4.2.1, 5.4.22, and the Section 5

references to allow the use of fuel enriched to 4.95 plus 0.05 weight

percent (w/o) U235.

The proposed license change is required to support delivery of

reload batch enrichments anticipated for Cycle 17 and beyond. These

reloads will require the use of fuel enrichments exceeding the current

TS limit of 4.20 plus 0.05 weight percent (w/o) U235 (nominal

4.20).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Increasing the allowable U235 enrichment

will have no influence on the probability of an accident previously

evaluated. No changes will be made to any safety related equipment,

systems, or setpoints used in determining the probability of an

evaluated accident. Neither will the proposed amendment allow

operation of the facility or safety equipment outside applicable

limitations or restrictions. Plant design bases will not be altered.

With respect to the Fuel Handling Accident, the manner in which the

fuel is handled will not be altered. The heat load on the Spent Fuel

Pool will not be increased and the cooling and circulation systems

and equipment will be unaltered. Therefore, there will be no

significant increase in the probability of an accident previously

evaluated.

The proposed change does not increase maximum allowable burnup

or fission product inventory. Since fission product inventory is an

inconsequential function of enrichment, radiological consequences

evaluated in the Updated Final Safety Analysis Report (UFSAR) will

not increase. The proposed change will not alter the function of

safety related equipment designed to mitigate the consequences of an

accident previously evaluated or allow operation of the facility

outside applicable limitations or restrictions. Accordingly the

proposed change will not involve a significant increase in the

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. The proposed increase in allowable enrichment will not

result in any design, operation, or function changes to any safety

related equipment designed to prevent and/or mitigate accidents, to

any setpoints or systems, or to any portion of the plant design

basis. Operation of the facility will remain within all required

limitations and restrictions. With respect to the Fuel Handling

Accident, the manner in which the fuel is handled will not be

altered. The heat load on the Spent Fuel Pool will not be increased

and the cooling and circulation systems and equipment will be

unaltered. Therefore, the proposed changes do not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety. NRC acceptance criteria and thus

the acceptable margin of safety to criticality for the Spent Fuel

Pool and New Fuel Storage Vault criticality are defined in Section

5.0 of the Technical Specifications. For the Spent Fuel Pool the

criteria specify that Keff must be maintained less than 0.95 when

the pit is flooded with unborated water. For the New Fuel Storage

Vault, the Keff must remain less than 0.95 if the vault is flooded

with unborated water, and must remain below 0.98 in an optimum

moderation event. Analyses performed in support of the proposed

change demonstrate that these acceptance criteria will continue to

be met. With respect to radiological consequences, the margin of

safety is defined by 10 CFR [Part] 100 limits which will not be

challenged. The analyses conclude that fission product inventory and

thus radiological consequences reported in Chapter 15 of the UFSAR

will not change. Accordingly the proposed license amendment will not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: July 29, 1994

Description of amendment request: The amendment would allow

operation of the plant with one Emergency Diesel Generator (EDG)

inoperable without entering a condition prohibited by Section 3.0 of

the Technical Specifications (TS). This TS request includes provisions

to avoid testing the operable EDG altogether under certain conditions

to ensure that one EDG is available to provide emergency power, if

needed, and to preserve the EDG overall life and reliability.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The proposed change involves changes in the

testing frequency of the EDGs when one EDG is inoperable, as well as

provision of additional measures to ensure that a source of off-site

power is available. The proposed change will also avoid testing of

an EDG when one EDG is inoperable if the EDG became inoperable for

reasons other than a common cause. Since the changes involve the

EDGs which perform an accident mitigation function and are not

involved in any accident initiation sequence, there is no

significant increase in the probability of a previously analyzed

accident. Since the changes involve the EDGs which perform an

accident mitigation function, and the changes provide additional

assurance that emergency power will be available for accident

mitigation, [there] is no significant increase in the consequences

of a previously analyzed accident. Therefore, there would be no

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. The proposed change involves changes in the testing

frequency of the EDGs when one EDG is inoperable, as well as

provision of additional measures to ensure that a source of off-site

power is available. The proposed change will also avoid testing of

an EDG when one EDG is inoperable if the EDG became inoperable for

reasons other than a common cause. Since these changes do not

involve changes in the operation of the plant, or physical or

equipment changes and involve controls for accident mitigation

equipment, the proposed amendment will not created the possibility

of new or different kind of accident from any accident previously

evaluated. Therefore, the proposed changes do not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety. The proposed change involves

changes in the testing frequency of the EDGs when one EDG is

inoperable, as well as provision of additional measures to ensure

that a source of off-site power is available. The proposed change

will also avoid testing of an EDG when one EDG is inoperable if the

EDG became inoperable for reasons other than a common cause. The

change reduces the required testing frequency of an operable EDG,

hence reducing time that no EDG will be available for automatic

starting and loading. These changes will provide assurance that

emergency power will be available to mitigate the effects of any

accident and will prevent excessive wear on the EDGs. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

NRC Project Director: David B. Matthews

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: July 22, 1993

Description of amendment request: The proposed amendment would

allow implementation of a performance based assessment program and the

corresponding functional and organizational changes in the Nuclear

Assessment Department. The changes affect the independent review

function, the independent assessment of plant activity, and the

independent Safety Engineering Group. These functions will be performed

by the proposed Nuclear Assessment Section.

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 change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated because it is a programmatic and administrative

change which does not physically alter any safety-related systems,

nor does it affect the way in which any safety-related systems

perform their functions. Since the design of the facility and system

operating parameters are not changing, the proposed amendment does

not involve an increase in the probability or consequences of any

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. As stated in Item 1, the proposed amendment is a

programmatic and administrative change which does not physically

alter any safety-related systems; nor does it affect the way in

which any safety-related systems perform their functions. Since the

design of the facility and system operating parameters are not

changing, the proposed amendment does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.3. The proposed amendment does not involve a significant

reduction in the

margin of safety because it is a programmatic and

administrative change which provides assurance that plant operations

continue to be conducted in a safe manner through the performance

based assessment programs. As stated in Item 1, the proposed

amendment does not physically alter any safety-related systems; nor

does it affect the way in which any safety-related systems perform

their functions. Since the design of the facility and system

operating parameters are not changing, the proposed amendment does

not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

NRC Project Director: David B. Matthews

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

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of amendment request: August 1, 1994

Description of amendment request: The proposed amendment would

revise the technical specifications to incorporate a 1.0 volt steam

generator tube interim plugging criteria (IPC) for Unit 1 beginning

with Cycle 7, which will begin in the fall of 1994.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Consistent with Regulatory Guide (RG) 1.121, ``Basis for

Plugging Degraded PWR Steam Generator Tubes, '' Revision 0, August

1976, the traditional depth-based criteria for SG tube repair

implicitly ensures that tubes accepted for continued service will

retain adequate structural and leakage integrity during normal

operating, transient, and postulated accident conditions. It is

recognized that defects in tubes permitted to remain in service,

especially cracks, occasionally grow entirely through-wall and

develop small leaks. Limits on allowable primary-to-secondary

leakage established in Technical Specifications ensure timely plant

shutdown before the structural and leakage integrity of the affected

tube is challenged.

The proposed license amendment request to implement voltage

amplitude SG tube support plate Interim Plugging Criteria for Byron

Unit 1 meets the requirements of RG 1.121. The IPC methodology

demonstrates that tube leakage is acceptably low and tube burst is a

highly improbable event during either normal operation or the most

limiting accident condition, a postulated main steam line break

(MSLB) event.

Adequate SG tube leakage integrity during normal operating

conditions is assured by limiting allowable primary-to-secondary

leakage to 150 gpd per SG or 600 gpd total. Currently, this limit is

administratively controlled. However, a license amendment request

was submitted on 06/03/94 to incorporate this limit into the Byron

Technical Specifications. During normal operating conditions, the

tube support plate constrains the [outer diameter stress corrosion

cracking] ODSCC affected area of the tube to provide additional

strength that precludes burst. Any leakage of a tube exhibiting

ODSCC at the [tube support plate] TSP is fully bounded by the

existing SG tube rupture analysis included in the Byron UFSAR.

Therefore, probability of failure of a tube left in service or

consequences of tube failure during normal operating conditions is

not significantly increased by the application of IPC.

During transients, the TSP is conservatively assumed to displace

due to the thermal-hydraulic loads associated with the transient.

This may partially expose a crack which is within the boundary of

the TSP during normal operations to free span conditions. Burst is

therefore conservatively evaluated assuming the crack is fully

exposed to free span conditions. The structural eddy current bobbin

coil voltage limit for free-span burst is 4.54 volts. This limit

takes into consideration a 1.43 safety factor applied to the steam

line break differential pressure that is consistent with RG 1.121

requirements. With additional considerations for growth rate

assumptions and an upper 95% confidence estimate on voltage

variability, the maximum voltage indication that could remain in

service is reduced to 2.7 volts. For added conservatism, the

allowable indication voltage is further reduced in the proposed

amendment to a 1.0 volt confirmed ODSCC indication limit. All

indications between 1.0 and 2.7 volts will be subject to an RPC

examination. Tubes with RPC confirmed ODSCC indications will be

plugged or sleeved. Any ODSCC indications between 1.0 volt and 2.7

volts which are not confirmed as ODSCC will be allowed to remain in

service since these indications are not as likely to affect tube

structural integrity or leakage integrity over the next operating

cycle as the indications that are detectable by both bobbin and

[rotating pancake coil] RPC inspections.

The eddy current inspection process has been enhanced to address

RG 1.83, ``Inservice Inspection of PWR Steam Generator Tubes,''

Revision 1, July 1975, considerations as well as the EPRI SG

Inspection Guidelines. Enhancements in accordance with NUREG-1477

and Appendix A of the Catawba IPC report (WCAP-13698) are in place

to increase detection of ODSCC indications and to ensure reliable,

consistent acquisition and analysis of data. Based on the

conservative selection of the voltage criteria and the increased

ability to identify ODSCC, the probability of tube failure during an

accident is also not significantly increased due to application of

requested IPC.

For consistency with current offsite dose limits, the site

allowable leakage limit during a MSLB has been conservatively

calculated to be 12.8 gpm. This leakage limit includes maximum

allowable operational leakage from the unaffected SGs and the

accident leakage from the affected SG. As a requirement for

operation following application of IPC, the projected distribution

of crack indications over the operating period must be verified to

result in primary to secondary accident leakage less than the site

allowable leakage limit. Thus, the consequences of a MSLB remain

unchanged.

Therefore, as implementation of the 1.0 volt IPC for Byron Unit

1 does not adversely affect steam generator tube integrity and

results in acceptable dose consequences, the proposed license

amendment request does not result in any significant increase in the

probability or consequences of an accident previously evaluated

within the Byron Updated Final Safety Analysis Report.

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

or different kind of accident from any accident previously

evaluated.

Implementation of the proposed SG tube IPC does not introduce

any significant changes to the plant design basis. Use of the

criteria does not provide a mechanism which could result in an

accident outside the tube support plate elevations since industry

experience indicates that ODSCC originating within the tube support

plate does not extend significantly beyond the thickness of the

support plate. This criteria only applies to ODSCC contained within

the region of the tube bounded by the tube support plate.

In addressing the combined effects of Loss of Coolant Accident

(LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG

(as required by General Design Criteria 2), it has been determined

that tube collapse of select tubes may occur in the SGs at some

plants, including Byron Unit 1. There are two issues associated with

SG tube collapse. First, the collapse of SG tubing reduces the RCS

flow area through the tubes. The reduction in flow area increases

the resistance to flow of steam from the core during a LOCA which,

in turn, may potentially increase Peak Clad Temperature (PCT).

Second, there is a potential that partial through-wall cracks in

tubes could progress to through-wall cracks during tube deformation

or collapse.

A number of tubes have been identified, in the ``wedge''

locations of the SG TSPs, that demonstrate the potential for tube

collapse during a LOCA + SSE event. Because of this potential, these

tubes have been excluded from application of the voltage-based SG

TSP IPC.

Therefore, neither a single or multiple tube rupture event would

be expected in a steam generator in which IPC has been applied.

ComEd has implemented a maximum primary to secondary leakage

limit of 150 gpd through any one SG at Byron to help preclude the

potential for excessive leakage during all plant conditions. The 150

gpd limit provides for leakage detection and plant shutdown in the

event of an unexpected single crack leak associated with the longest

permissible free span crack length. The 150 gpd limit provides

adequate leakage detection and plant shutdown criteria in the event

an unexpected single crack results in leakage that is associated

with the longest permissible free span crack length. Since tube

burst is precluded during normal operation due to the proximity of

the TSP to the tube and the potential exists for the crevice to

become uncovered during MSLB conditions, the leakage from the

maximum permissible crack must preclude tube burst at MSLB

conditions. Thus, the 150 gpd limit provides a conservative limit to

prompt plant shutdown prior to reaching critical crack lengths under

MSLB conditions.

Upon implementation of the 1.0 volt IPC, steam generator tube

integrity continues to be maintained through inservice inspection

and primary-to-secondary leakage monitoring. Therefore, the

possibility of a new or different kind of accident from any

previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The use of the voltage based bobbin coil probe SG TSP IPC for

Byron Unit 1 will maintain steam generator tube integrity

commensurate with the criteria of RG 1.121 as discussed above. Upon

implementation of the criteria, even under the worst case

conditions, the occurrence of ODSCC at the TSP elevations is not

expected to lead to a steam generator tube rupture event during

normal or faulted plant conditions. The distribution of crack

indications at the TSP elevations result in acceptable primary-to-

secondary leakage during all plant conditions and radiological

consequences are not adversely impacted by the application of IPC.

The installation of SG tube plugs and sleeves reduces the RCS

flow margin. As noted previously, implementation of the SG TSP IPC

will decrease the number of tubes which must be repaired by plugging

or sleeving. Thus, implementation of IPC will retain additional flow

margin that would otherwise be reduced due to increased tube

plugging. Therefore, no significant reduction in the margin of

safety will occur as a result of the implementation of this proposed

license amendment request.

Although not relied upon to prove adequacy of the proposed

amendment request, the following analyses demonstrate that

significant conservatisms exist in the methods and justifications

described above:

LIMITED TUBE SUPPORT PLATE DISPLACEMENT

An analysis was performed to verify [the effect] of limited TSP

displacement during accident conditions (MSLB). Application of

minimum TSP displacement assumptions reduce the likelihood of a tube

burst to negligible levels. Consideration of limited TSP

displacement would also reduce potential MSLB leakage when compared

to the leakage calculated assuming free span indications.

PROBABILITY OF DETECTION

The Electric Power Research Institute (EPRI) Performance

Demonstration Program analyzed the performance of approximately 20

eddy current data analysts evaluating data from a unit with 3/4''

inside diameter and 0.049'' wall thickness tubes. The results of

this analysis clearly show that the detectability of larger voltage

indications is increased which lends creditability for application

of a POD of 0.62 for ODSCC indications larger than 1.0

volt.

RISK EVALUATION OF CORE DAMAGE

As part of ComEd's evaluation of the operability of Byron Unit 1

Cycle 7, a risk evaluation was completed. The objective of this

evaluation was to compare core damage frequency under containment

bypass conditions, with and without the interim plugging criteria

applied at Byron Unit 1.

The total Byron core damage frequency is estimated to be 3.09E-5

per reactor year with a total contribution from containment bypass

sequences of 3.72E-8 per reactor year according to the results of

the current individual plant evaluation (IPE). Operation with the

requested IPC resulted in an insignificant increase in core damage

frequency resulting from MSLB with containment bypass conditions.

Therefore, based on the evaluation above, ComEd has concluded

that this proposed license amendment request does not involve a

significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Byron Public Library, 109 N.

Franklin, P.O. Box 434, Byron, Illinois 61010

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County,

Illinois;Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power

Station, Units 1 and 2, Rock Island County, Illinois;Docket Nos.

50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 2, Lake

County, Illinois

Date of amendment request: July 8, 1994

Description of amendment request: The proposed amendment would add

a License Condition to specify that commitments made in response to the

March 14, 1983, NUREG-0737 Order shall be maintained pursuant to the

requirements of 10 CFR 50.59.

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:

Involve a significant increase in the probability or

consequences of any accident previously analyzed:

Commonwealth Edison has addressed all issues made in response to

NUREG-0737. As such, the purpose of the post-TMI Order is no longer

served. The inclusion of the modified Order as a license condition

is administrative in nature and does not allow unregulated decreases

in the level of safety; therefore, this license amendment is

appropriate and safe. The proposed license amendment requires

control of NUREG-0737 commitments through 10 CFR 50.59. If an

unreviewed safety question occurs during the review of a NUREG 0737

item then Commonwealth Edison is obligated to submit a change to the

NRC staff as a license amendment. As a result of the proposed

amendment, there are no physical changes to the facility and all

operating procedures, limiting conditions for operation (LCO),

limiting safety system settings, and safety limits specified in the

Technical Specifications will remain unchanged. Therefore, the

proposed license amendment to modify the post-TMI Order will not

increase the probability or the consequences of any accident

previously analyzed.

Create the possibility of a new or different kind of accident

from any previously evaluated:

Since there are no changes in the way the plant is operated, the

potential for a new or different kind of accident is not created.

The proposed changes are administrative in nature and do not affect

any accident initiators for Dresden, Quad Cities, and Zion Stations.

No new failure modes are introduced.

Involve a significant reduction in a margin of safety:

Plant safety margins are established through LCOs, limiting

safety system settings, and safety limits specified in the Technical

Specifications. As a result of the proposed amendment, there will be

no changes to either the physical design of the plant or to any of

these settings and limits. The proposed changes are administrative

and do not affect the safe operation of the sites. Therefore, there

will be no changes to any of the margins of safety.

Guidance has been provided in 51 FR 7744 for the application of

standards to license change requests for determination of the

existence of significant hazards considerations. This document

provides examples of amendments which are not likely considered to

involve significant hazards considerations.

This proposed amendment does not involve a significant

relaxation of the criteria used to establish safety limits, a

significant relaxation of the bases for the limiting safety system

setting or a significant relaxation of the bases for the limiting

conditions for operations. The proposed changes are administrative

in nature without consequence to the safety of the plant. Therefore,

based on the guidance provided in the Federal Register and the

criteria established in 10 CFR 50.92(c), the proposed change does

not constitute a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,

Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021; for

Zion, Waukegan Public Library, 128 N. County Street, Waukegan, Illinois

60085

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company, and Northeast Nuclear

Energy Company, Docket Nos. 50-213, 50-245, 50-336, and 50-423

Haddam Neck Plant, and Millstone Nuclear Power Station, Units 1, 2,

and 3, Middlesex County, and New London County, Connecticut

Date of amendment request: June 30, 1994

Description of amendment request: The proposed amendments would

modify the Administrative Controls Section of the Technical

Specifications by replacing the present Nuclear Review Board (NRB) for

the Haddam Neck Plant, and the NRB and Site Nuclear Review Board (SNRB)

with a Nuclear Safety Assessment Board (NSAB) which will serve

Millstone Units 1, 2, and 3, and Haddam Neck.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

... These proposed changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The NSAB is an oversight group which provides independent

assessments of activities at the Haddam Neck Plant and Millstone

Unit Nos. 1, 2, and 3. The members of the NSAB are appointed by the

Executive Vice President - Nuclear to provide oversight and feedback

on the operation of the units. The NSAB adds to the defense-in-depth

provided by the design, operation, maintenance, and quality

oversight of the nuclear units by promoting excellence through the

conduct of its affairs and advising the Executive Vice President -

Nuclear in matters concerning nuclear safety.

The proposed modification to the Technical Specifications are

administrative in nature and will establish a new group which will

accomplish the guidance provided in ANSI N18.7-1976. The charter of

the NSAB will be controlled by procedure.

These administrative changes will not increase the probability

of occurrence or the consequences of an accident previously

evaluated.

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

from any previously evaluated.

The proposed addition of the NSAB and its subcommittees and the

ensuing elimination of the NRB and the SNRB is an administrative

reorganization. There are no changes in the way in which the plants

are physically operated. The administrative changes being

accomplished by the establishment of the NSAB fulfills the function

previously provided by the NRB and the SNRB. The organization of the

NSAB will follow the guidance found in ANSI N18.7-1976 and will be

controlled by procedure.

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

The proposed changes establish the requirements of the NSAB. The

NSAB replaces those activities previously performed by the NRB and

the SNRB. With these changes the new organization will provide more

consistent and clearer feedback to the four units and the Executive

Vice President - Nuclear.

The changes do not directly affect any protective boundaries nor

do they impact the safety limits for the protective boundaries.

These proposed changes are administrative in nature. Therefore,

there can be 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: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457, for the Haddam Neck Plant, and

the Learning Resource Center, Three Rivers Community-Technical College,

Thames Valley Campus, 574 New London Turnpike, Norwich, Connecticut

06360, for Millstone 1, 2 and 3.

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

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

Connecticut, 06141-0270.

NRC Project Director: John F. Stolz

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: August 11, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification Section 6.5.1, Station Nuclear Safety

Committee (SNSC), to change the designation of the Chairman and to

clarify the maximum allowable alternate members for quorum purposes.

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.

This is an administrative change. Since this change maintains a

consistent level of chairmanship while continuing to ensure

independence and technical expertise of the SNSC chairman, this

change does not increase the probability or consequences of an

accident.

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

any previously evaluated has not been created.

This is an administrative change of the designation of the

Chairman of SNSC which does not significantly decrease the level of

senior management which is responsible for chairing SNSC. No new or

different kind of accident has been created.

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

The independence and technical expertise of the SNSC Chairman

will be preserved. SNSC will continue to be composed of those

individuals most related to matters of nuclear safety. The margin of

safety will not be reduced by this change.

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: Pao Tsin Kuo

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

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

Date of amendment request: 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 4.7.13.3 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 administrative 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February 5, 1993 as supplemented by

letter dated August 1, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to incorporate a technical review

and control process to supplement the onsite technical review and

approval of new procedures and changes thereto affecting nuclear

safety. This process is discussed in Section 5.5 of the Revised

Standard Technical Specifications, NUREG-1432. This notice supersedes

the notice issued on April 14, 1993 (58 FR 19478), and acknowledges the

clarification in the licensee's August 1, 1994, letter.

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 is administrative in nature and provides for

1) procedural reviews through the use of qualified technical review

personnel designated by the PORC [Plant Operating Review Committee]

and 2) procedural approval through the use of group heads designated

by the General Manager Plant Operations as authorized by

administrative controls upon their development. As part of this

process, qualified technical reviewers will be individuals other

than the preparer who will document and implement necessary cross-

discipline reviews prior to approval. The process will be controlled

by administrative controls which will be reviewed by the PORC and

approved by the General Manager Plant Operations.

The procedures governing plant operation will continue to ensure

that plant parameters are maintained within acceptable limits.

Procedures and changes thereto will be reviewed and approved at a

level commensurate with their importance to safety. Therefore, the

proposed changes will not involve a significant increase in the

probability or consequences of any accident previously evaluated.

The proposed changes are administrative in nature. The proposed

changes do not involve physical changes to the plant, changes to

setpoints, or operating parameters. The applicable procedures

governing the operation of the plant will receive reviews and

approvals commensurate with their importance to nuclear safety, and

where appropriate cross-discipline review will be performed.

Therefore, the proposed changes will not create the possibility of a

new or different kind of accident from any previously evaluated.

The proposed changes are administrative in nature. The Waterford

3 safety margins are defined and maintained by the Technical

Specifications in Sections 2-5 which are unaffected. Therefore, the

proposed change will not involve a significant reduction in a margin

of safety.

The licensee's letter dated August 1, 1994, provided a

clarification of the proposed wording of the technical specifications

to assure the personnel performing the technical reviews would have the

necessary technical knowledge base.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August 19, 1994

Description of amendment request: The proposed amendment would move

the requirements of Technical Specification 3/4.3.4 Turbine Overspeed

Protection from the technical specifications (TS) and relocate them in

the Updated Final Safety Analysis Report (UFSAR) consistent with the

NRC Final Policy Statement on Technical Specifications Improvements for

Nuclear Power Reactors.

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 relocates the Turbine Valve Overspeed

Protection requirements from the TS to the Waterford 3 UFSAR

consistent with the NRC Policy Statement on Technical Specification

Improvements. Testing and inspections of the turbine Overspeed

Protection System will remain governed by an approved turbine

maintenance program, described in the UFSAR. This proposed change

has no affect on the current Turbine Overspeed Protection

requirements other then to relocate them to the UFSAR. Thus, the

probability of a turbine missile causing damage to a safety-related

component or structure at Waterford 3 as described in the FSAR

analysis (Reference 5) is not affected. The purpose of the Turbine

Overspeed Protection System is to prevent an overspeed event, the

precursor to a potential turbine fragment missile. Since the purpose

of this system is preventive, it serves no function to mitigate any

accident previously evaluated.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed change does not involve any change to the

configuration or method of operation of any plant equipment. No new

failure modes or limiting failures have been identified as result of

the proposed change. The proposed change will not alte the operation

of the plant or the manner in which it is operated. Any subsequent

change to the Turbine Oversspeed Protection System requirements will

undergo a review in accordance with the criteria of 10 CFR 50.59 to

ensure that the change does not involve an unreviewed safety

question.

Therefore, the proposed change will not create the possibility

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

evaluated.

The proposed change will relocate Turbine Overspeed Protection

System requirements from the TS to the Waterford 3 UFSAR on the

basis that the Turbine Overspeed Protection System does not meet the

criteria of the NRC Final Policy Statement on Technical

Specifications Improvements for Nuclear Reactors. The requirements

that will reside in the UFSAR for the Turbine Overspeed Protection

system will ensure that the system remains capable of protecting the

turbine from excessive overspeed. The proposed change will have no

adverse impact on any protective boundary or safety limit.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

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

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

Date of amendment request: July 25, 1994

Description of amendment request: The proposed amendments evise

various Technical Specification sections to implement enhancements

recommended by NRC Generic Letter (GL) 93-05, ``Line-Item Technical

Specification Improvements to Reduce Surveillance Requirements for

Testing During Power Operation,'' for St. Lucie Units 1 and 2.

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 standards used to arrive at a determination that a request

for amendment involves a no significant hazards consideration are

included in the Commission's regulation, 10 CFR 50.92. 10 CFR 50.92

states that no significant hazards considerations are involved if

the 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. Each standard is discussed as

follows:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed amendments conform to the guidance given in

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

capabilities of the incore detector system, reactor coolant system

pressure isolation valves, safety injection tank, or containment

sump will not be modified by the proposed change. Therefore, the

probability or consequences of an accident are not significantly

increased by the changes.

(2) Use of the modified specification would not create the

possibility of a new or different kind of accident from any

previously evaluated.

The use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

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

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

the surveillance interval changes and clarifications, since the

proposed changes do not involve the addition or modification of

equipment nor do they alter the design or operation of affected

plant systems.

(3) Use of the modified specification would not involve a

significant reduction in a margin of safety.

The operating limits and functional capabilities of the affected

systems are unchanged by the proposed amendments. Therefore, the

modified specifications which establish new or clarify old

surveillance intervals consistent with the NRC Generic Letter 93-05

line-item improvement guidance do not significantly reduce any of

the margins of safety.

Based on the above, we have determined that the proposed

amendments do not (1) involve a significant increase in the

probability or consequences of an accident previously evaluated, (2)

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

any previously evaluated, or (3) involve a significant reduction in

a margin of safety; and therefore do not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Project Director: Victor McCree, Acting

Florida Power and Light Company, et al., Docket No. 50-389, St.

Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: July 25, 1994

Description of amendment request: The amendment will upgrade

Technical Specification (TS) 3/4.7.1.6 for the Main Feedwater Line

Isolation Valves to be consistent with NUREG-1432, ``Standard Technical

Specifications for Combustion Engineering Plants.'' The changes include

all related requirements of NUREG-1432, Revision O, specification

3.7.3. Accordingly, the proposal is consistent with the Commission's

Final Policy Statement on Technical Specifications Improvements (58 FR

39132).

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:

Pursuant to 10 CFR 50.92, a determination may be made that a

proposed license amendment involves no significant hazards

consideration if 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. Each standard is discussed as

follows:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment will upgrade the existing Limiting

Condition for Operation (LCO) associated with the Main Feedwater

Line Isolation Valves (MFIVs) to be consistent with NUREG-1432,

Standard Technical Specifications for Combustion Engineering Plants.

The MFIVs are not initiators of accidents previously evaluated, but

are included as part of the success paths associated with mitigating

various accidents and transients. The redundancy afforded by two

MFIVs per feedwater line in conjunction with the requirements of the

proposed LCO assure that the feedwater isolation safety function of

these valves can be accomplished considering single failure

criteria. Neither the feedwater system design nor the safety

function of the MFIVs have been altered from those previously

evaluated, and the proposed amendment does not change the applicable

plant safety analyses.

Therefore, operation of the facility in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the

modes of operation defined in the facility license. The changes are

administrative in nature in that they do not involve the addition of

new equipment or the modification of existing equipment, nor do they

otherwise alter the design of St. Lucie Unit 2 systems. Therefore,

operation of the facility in accordance with the proposed amendment

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

accident from any accident previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The safety function of the MFIVs is to terminate main feedwater

flow and isolate the safety related portion from the non-safety

related portion of the feedwater system. The proposed amendment, in

conjunction with the redundancy afforded by the feedwater system

design, assures that this safety function can be accomplished

considering single-failure criteria. The bases for required actions

and the action completion times specified for inoperable MFIVs is

consistent with the corresponding specifications in NUREG-1432,

which are equally applicable to St. Lucie Unit 2. The safety

analyses for applicable accidents and transients remain unchanged

from those previously evaluated and reported in the Updated Final

Safety Analysis Report. Therefore, operation of the facility in

accordance with the proposed amendment would not involve a

significant reduction in a margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS Changes, FPL has concluded that this

proposed license amendment involves no significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Project Director: Victor McCree, Acting

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: July 18, 1994

Description of amendment request: The licensee proposes to revise

Technical Specification Table 4.3-1, Reactor Trip System

Instrumentation Surveillance Requirements, Technical Specification

3.3.4, Turbine Governor Valves and Technical Specification 3.7.1.2,

Turbine Driven Auxiliary Feedwater Pump. The purpose of this amendment

is to remove one-time amendments that are no longer necessary. In

addition, six minor editorial changes are proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of a previously evaluated

accident.

The changes proposed to remove the one-time amendments return

the Technical Specifications to the exact wording prior to the one-

time amendments. Returning the Technical Specifications to their

original wording is administrative because the one-time amendments

are no longer applicable. Hence, removing the one-time amendments

would not increase the probability or consequences of an accident.

The other changes are purely editorial in nature, hence, would not

increase the probability or consequences of an accident. Based on

the above, removal of the one-time amendments from the Technical

Specifications will not significantly increase the probability or

consequences of an accident.

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

or different kind of accident from any previously evaluated.

The changes proposed to remove the one-time amendments return

the Technical Specifications to the exact wording prior to the one-

time amendments. Returning the Technical Specifications to their

original wording is administrative because the one-time amendments

are no longer applicable. Therefore, removing the one-time

amendments would not create the possibility of a new or different

kind of accident. The other changes are purely editorial in nature,

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

of accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The changes proposed to remove the one-time amendments return

the Technical Specifications to the exact wording prior to the one-

time amendments. Returning the Technical Specifications to their

original wording is administrative because the one-time amendments

are no longer applicable. Therefore, removing the one-time

amendments would not involve a significant reduction in a margin to

safety. The other changes are purely editorial in nature, hence,

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

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior

College, J. M. Hodges, Learning Center, 911 Boling Highway, Wharton,

Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, N.W., Washington, D.C. 20036

NRC Project Director: William D. Beckner

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

Linn County, Iowa

Date of amendment request: July 12, 1994

Description of amendment request: The proposed amendment changes

the requirement to perform the surveillance test for the channel

functional test Rod Block Monitor, Flow-biased Average Power Range

Monitor and Recirculation Flow instruments from within 24 hours prior

to startup to after the reactor is in the RUN mode, but prior to when

each system is assumed to function in the plant safety analysis.

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 to the Channel Functional Test frequency

for the RBM will not significantly increase the probability or

consequences for any previously-evaluated event as we are only

matching the mode requirements for performing the SR to the

OPERABILITY requirement for the RBM system, i.e., prior to 30% RTP.

The system will be verified to be OPERABLE prior to when it is

assumed to be OPERABLE in the Updated Final Safety Analysis Report

(UFSAR) for the DAEC.

Allowing the Channel Functional Test for the APRM Flow-Biased

Rod Block Upscale and Downscale trips to be performed ``within 24

hours of entering RUN mode and prior to exceeding 25% RTP'' will not

increase either the probability or consequences of any previously-

analyzed event. The applicable event for the rod block function

during reactor startup is a control Rod Withdrawal Error (RWE),

which is initiated by either an operator error or malfunction within

the Reactor Manual Control System, not by a malfunction within the

APRM system. However, a RWE event that could challenge the fuel

thermal limits is precluded because, as documented in the DAEC UFSAR

(see Section 15.4) and the analysis submitted to support DAEC TS

Amendment No. 120,( NEDC-30813-P,

Average Power Range Monitor, Rod Block Monitor and Technical

Specification Improvement (ARTS) Program for the Duane Arnold Energy

Center, December 1984.), significant margin exists below 25% RTP to

assure the Safety Limit Minimum Critical Power Ratio (SLMCPR) is not

violated by a RWE event. In addition, rod pattern controls are in

place during this period to limit the rod withdrawal sequence, i.e.,

rod worth, such that the fuel thermal limits would not be exceeded.

The Control Rod Drop Accident is unaffected by the requested SR

change as the ``accident'' control rod is assumed to be de-coupled

from its drive mechanism and free-falls from fully inserted to

fully-withdrawal. As the drive for that rod is assumed to be fully-

withdrawn as an initial condition in the event, the APRM rod block

has no role in either preventing or mitigating the rod drop

accident. Thus, revising the SR for the APRM Flow-Biased Rod Block

has no impact upon the Control Rod Drop Accident.

The SRs for the Recirculation Flow Rod Block trips are being

modified for consistency with the APRM Rod Block changes above, as

the sole purpose of this Recirculation Flow signal is to provide the

flow input signal into the APRM Flow-Biased trips. The Recirculation

Flow units are a support system to the APRM Flow-Biased Rod Blocks.

There is no event that is either caused by or mitigated by the

Recirculation Flow Rod Block trips. They are provided solely to

ensure that if the flow signal being input into the APRM circuits is

not valid, a precautionary rod block will be generated as the APRM

Flow-Biased Rod Block setpoint could be in error. Consequently,

allowing the Channel Functional Test for the Recirculation Flow Rod

Block Upscale, Downscale and Comparator trips to be performed

``within 24 hours of entering RUN mode and prior to exceeding 25%

RTP'' will not increase either the probability or consequences of

any previously-analyzed event as these rod blocks are not involved

in either preventing or mitigating any analyzed event.

2) The proposed change to the Channel Functional Test frequency

for the RBM will not introduce any new or different event, as no

changes in system design or operation are being made. We are only

matching the requirement for performing the SR to the OPERABILITY

requirement for the RBM system.

The proposed change to the Channel Functional Test frequency for

the APRM and Recirculation Flow Rod Blocks will not introduce any

new or different event, as no changes in either system design or

operation are being made. In fact, by allowing the Channel

Functional Test to be performed in an operating state which does not

require extensive use of jumpers and/or relay blocks, we reduce the

possibility of an error being made that could cause an inadvertent

actuation of an ESF or disabling of an ESF.

3) The proposed change matches the mode requirement for

performing the SR to the OPERABILITY requirement for the RBM system,

i.e., prior to 30% RTP. The system will be verified to be

OPERABLE prior to when it is assumed to be OPERABLE in the UFSAR

accident analysis. Thus, the margin of safety for the RBM is not

reduced.

As stated in the BASES for TS Chapter 3/4.2, the margin of

safety for the APRM rod block is to prevent violation of the SLMCPR

in RUN by a RWE event. The analysis of the RWE event during Startup

(See DAEC UFSAR Section 15.4.2) and during Power Operation (Ibid),

demonstrates that violations of the SLMCPR are not possible in RUN

below 25% RTP when normal control rod patterns are followed (which

are reinforced by procedural and/or automatic rod pattern controls).

Because the proposed change to the SR for the APRM Flow-Biased Rod

Block will still ensure that the trip will be OPERABLE prior to

exceeding 25% RTP, this change will not reduce the existing margin

of safety.

Again, the Recirculation Flow units are a support system to the

APRM flow-biased circuits. The Recirculation Flow Rod Blocks are

merely precautionary, they do not prevent or mitigate any accident.

Therefore, the proposed revision to the Recirculation Flow Rod block

SR frequency will not reduce the margin of safety for the same

reasons given above for the APRM Rod Blocks.

The NRC staff 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: Cedar Rapids Public Library,

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

Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,

Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC

20036.

NRC Project Director: John N. HannonIES Utilities Inc., Docket No.

50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: July 29, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications by allowing the processing and

implementation of an ISI or IST request for relief from the ASME Code

under 10 CFR 50.59 without prior NRC approval, provided that the relief

request has been reviewed and approved by the plant staff and plant

safety committee.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Operation of the facility in accordance with the proposed

amendment would not involve any increase in the probability of

occurrence or consequences of an accident previously evaluated. The

Inservice Inspection and Testing Programs, pursuant to 10 CFR

50.55a, are described in the Technical Specifications. The proposed

amendment, in accordance with NUREG-1433 and draft NUREG-1482,

permits relief from an ASME Code requirement in the interim between

the time of submittal of a relief request and NRC approval of the

relief. The changes being proposed do not affect assumptions

contained in plant safety analyses or change the physical design

and/or operation of the plant, nor do they affect Technical

Specifications that preserve safety analysis assumptions. Any relief

from the approved ASME Section XI Code requirements that is

implemented prior to NRC review and approval will require evaluation

under the 10 CFR 50.59 process to determine that no TS changes or

unreviewed safety questions exist. This evaluation process will

ensure that the impact of any Code relief is thoroughly evaluated

and that the structures, systems and components remain in

conformance with assumptions made in the safety analysis. Therefore,

operation of the facility in accordance with the proposed amendment

would not affect the probability or consequences of an accident

previously evaluated.

2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated. The

Inservice Inspection and Testing Programs, pursuant to 10 CFR

50.55a, are described in the Technical Specifications. The proposed

amendment, in accordance with NUREG-1433 and draft NUREG-1482,

permits relief from an ASME Code requirement in the interim between

the time of submittal of a relief request and NRC approval of the

relief. The changes being proposed will not change the physical

plant or the modes of operation defined in the Facility License. The

changes do not involve the addition or modification of equipment nor

do they alter the design or operation of plant systems. Any relief

from the approved ASME Section XI Code requirements that is

implemented prior to NRC review and approval will require evaluation

under the 10 CFR 50.59 process to determine that no TS changes or

unreviewed safety questions exist. This evaluation process will

ensure that the impact on any Code relief is thoroughly evaluated

and that the structures, systems and components remain in

conformance with assumptions made in the safety analysis. Therefore,

operation of the facility in accordance with the proposed amendment

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

accident previously evaluated.

3) Operation of the facility in accordance with the proposed

amendment would not involve any reduction in a margin of safety. The

Inservice Inspection and Testing Programs, pursuant to 10 CFR

50.55a, are described in the Technical Specifications. The proposed

amendment, in accordance with NUREG-1433 and draft NUREG-1482,

permits relief from an ASME Code requirement in the interim between

the time of submittal of a relief request and NRC approval of the

relief. The changes being proposed do not alter the bases for

assurance that safety-related activities are performed correctly or

the basis for any TS that is related to the establishment of or

maintenance of a safety margin. Any relief from the approved ASME

Section XI Code equirements that is implemented prior to NRC review

and approval will require evaluation under the 10 CFR 50.59 process

to determine that no TS changes or unreviewed safety questions

exist. This evaluation process will ensure that the impact on any

Code relief does not affect the ability of structures, systems or

components to perform their design function, affect compliance with

any TS requirements or reduce the margin of safety. Therefore,

operation of the facility in accordance with the proposed amendment

would not involve a reduction in a margin of safety.

The NRC staff 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: Cedar Rapids Public Library,

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

Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,

Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC

20036

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: July 26, 1994

Description of amendment request: The proposed amendment would

revise the existing limiting condition for operation (LCO) 3.12.A.2.c

to allow for increased flow capacity of the control room emergency

filter system. By increasing the maximum allowed makeup capacity of

this system, additional margin is provided for the positive

pressurization of the control room envelope.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

Evaluation

This license amendment request involves the upgrading of the

Control Room Emergency Filter System from 341 cubic feet per minute

(CFM) plus or minus 10% to a maximum of Q(z) when using the base load option.

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 would 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 evaluated.

The proposed changes are clarifications or relocation of

existing technical specification requirements and do not

substantively affect plant operation. Since they do not affect plant

operations, they cannot be initiators of any events.

The safety analysis of the plant is unaffected by the proposed

changes. Since the safety analysis is unaffected, the calculated

radiological releases associated with the accident analyses are not

affected. Therefore, the proposed changes will not increase the

probability or consequences of previously evaluated accidents.

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

from any accident previously evaluated.

As previously stated, the proposed changes are clarifications or

relocation of existing technical specifications and do not

substantively affect plant operation. No new failure modes are

introduced. Since the proposed modifications do not affect plant

operations, they cannot be initiators of new events.

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

The proposed changes are clarifications or relocation of

existing technical specifications and are not substantive changes.

The correction of the title in Figure 3.1-5 will ensure consistency

throughout the technical specifications. The relocation of the CVCS

valves requirements from the RCS [Reactor Coolant System] - Cold

Shutdown Specification to the Reactivity Control Systems - Shutdown

Margin specification will ensure the CVCS valves requirements are

located in the most appropriate location and will help the operators

from the commission of errors or omission of actions due to

inappropriately located material. The final change will revise the

action statement sections of the specification pertaining to heat

flux hot channel factor to ensure all actions in these

specifications are clearly displayed and not contained in the

corresponding surveillance requirements. Therefore, since these

changes are editorial in nature, the proposed modification will have

no impact on the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

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

Connecticut 06141-0270.

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket No. 50-171, Peach Bottom

Atomic Power Station, Unit 1, York County, Pennsylvania

Date of Application for Amendment: May 9, 1994

Brief description of amendment: This Licensee Amendment Request

(LAR) proposes to revise the Peach Bottom Atomic Power Station, Unit 1,

Possession-Only License and Technical Specifications (TS) to reflect

the name change of Philadelphia Electric Company to PECO Energy

Company, to provide proper reference to 10 CFR Part 20 requirements (56

FR 23360), and to reduce the required frequency for performing periodic

inspections in the containment vessel below ground level for water

accumulation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. Does the proposed amendment involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed changes do not alter the operation of equipment

assumed to be an initiator or any analyzed event or assumed to be

available for the mitigation of accidents or transients. Proposed

changes 1 and 2 are administrative in nature. Proposed change 3 to

reduce the required frequency for performing the periodic inspection

for water accumulation in the containment vessel below ground level

does not impact the probability of ground water intrusion into the

containment building. Proposed change 3 maintains adequate assurance

that integrity of the containment building with respect to ground

water entry will be maintained. The design of Unit 1 makes it very

difficult for ground water to reach the exterior of the containment

liner to start the metal corrosion process. The concrete layer

between the rock and the containment liner serves as a barrier to

prevent water migration to the liner shell. A cathodic protection

system provides protective current to the containment liner as well

as nearby underground piping. The steel containment liner of Unit 1

should not corrode under the present environmental conditions or any

anticipated future conditions even without an operating cathodic

protection system. Monthly inspections from May 1990 (following

issuance of Amendment No. 7 to the Possession-Only License No. DR-12

on April 25, 1990) through April 1994 have not detected any water in

the containment building. Prior to Amendment No. 7, the inspection

of Unit 1 was performed semi-annually. A review of these semi-annual

inspections dating back to October 1981 determined that water has

never been detected in the accessible areas below ground level in

the containment building. The TS limit water accumulation in the

containment sump to 500 gallons. Twelve and one-half years of

inspections have confirmed the reliability of the design of Unit 1

to maintain integrity against any ground water intrusion. There is

no reason, based on the review of inspection data, why the

inspection could not be performed semi-annually rather than monthly.

Therefore, these proposed changes do not increase the probability or

consequences of an accident previously evaluated.

b. Does the proposed amendment create the possibility of a new

or different kind of accident from any accident previously

evaluated?

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

different kind of accident from any accident previously evaluated

because implementation of the proposed changes do not involve any

physical changes to plant systems, structures, or components. The

proposed changes do not affect the plant SAFSTOR status. Therefore,

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

accident previously evaluated is not created.

c. Does the proposed amendment involve a significant reduction

in a margin of safety?

The proposed changes do not involve a significant reduction in a

margin of safety because the proposed changes do not affect the

plant SAFSTOR status. Because proposed changes 1 and 2 are

administrative in nature, they do not involve a question of safety.

The semi-annual inspection of the accessible areas below ground

level in the containment building for water accumulation, as

proposed by change 3, is adequate to ensure containment building

integrity is maintained with respect to ground water. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105

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

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

Pennsylvania 19101

NRC Branch Chief: John H. Austin

Power Authority of the State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: August 4, 1994

Description of amendment request: The proposed changes would revise

Sections 3.4 and 3.5 of the Technical Specifications. The Section 3.4

revision would reduce the maximum allowable percent of rated power

associated with inoperable Main Steam Safety Valves (MSSVs). This

change would modify Table 3.4-1 and the associated Basis such that the

maximum power level allowed for operation with inoperable MSSVs is

below the heat removing capability of the operable MSSVs. The Section

3.5 revision would correct administrative errors in the action

statements associated with Items 2.a and 2.c of Table 3.5-4.

Additionally, the proposed changes to Item 2.b of Table 3.5-3 and Item

2.b of Table 3.5-4 would clarify the action statements associated with

inoperable high containment pressure (Hi-Hi Level) instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Consistent with the criteria of 10 CFR 50.92, the enclosed

application is judged to involve no significant hazards based on the

following information:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. This proposed technical specification change

would modify Table 3.4-1 and the associated basis such that the

maximum power level allowed for operation with inoperable MSSVs is

below the heat removing capability of the operable MSSVs. This

proposed technical specification change will be more conservative

than the current technical specifications. Proposed changes to Items

2.a and 2.c of Table 3.5-4 would restore the original intent of the

specifications and remove undue restrictions on the plant. Proposed

changes to Item 2.b of Table 3.5-3 and Item 2.b of Table 3.5-4

clarify the action statements associated with inoperable high

containment pressure (Hi-Hi Level) instrumentation.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed license amendment does not create the possibility

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

evaluated. The proposed change incorporates more conservative limits

on the maximum power level allowed for operation with inoperable

MSSVs, restores the original intent of items 2.a and 2.c of Table

3.5-4, and clarifies action statements associated with item 2.b of

Table 3.5-3 and item 2.b of Table 3.5-4.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed amendment would not involve a significant reduction

in a margin of safety. This proposed technical specification change

would modify Table 3.4-1 and the associated basis such that the

maximum power level allowed for operation with inoperable MSSVs is

below the heat removing capability of the operable MSSVs. This

proposed technical specification change will be more conservative

than the current technical specifications. Proposed changes to Items

2.a and 2.c of Table 3.5-4 would restore the original intent of the

specifications and remove undue restrictions on the plant. Proposed

changes to Item 2.b of Table 3.5-3 and Item 2.b of Table 3.5-3 and

Item 2.b of Table 3.5-4 clarify the action statements associated

with inoperable high containment pressure (Hi-Hi Level)

instrumentation.

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 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Pao Tsin Kuo

Power Authority of The State of New York, Docket No. 50-286,

Indian PointNuclear Generating Unit No. 3, Westchester County, New

York

Date of amendment request: August 4, 1994

Description of amendment request: The proposed changes would revise

the fuel oil availability requirements for the Emergency Diesel

Generators (EDGs) from Section 3.7 of the Technical Specifications

(TSs). This TS change would require that 30,026 gallons of fuel oil be

available onsite in addition to the oil in the EDG storage tanks. TS

3.7.F.4 is also being changed to require a total of 7056 gallons of

fuel in the EDG fuel oil storage tanks. In addition, several

administrative changes are being proposed to remove the word

``available'' from the phrase ''... gallons of fuel available...'' in

Section 3.7.A.5 (for the individual storage tanks) to avoid confusion

regarding the amount of usable fuel in the tanks.

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: Consistent with the criteria

of 10 CFR 50.92, the enclosed application is judged to involve no

significant hazards based on the following information:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response:

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously analyzed.

The change in the minimum required volume for the EDG fuel oil

storage tanks ensures that two EDGs can power minimum safeguards

equipment for 48 hours. The new required levels allow for

temperature effects on fuel density and calibration uncertainties.

The change to the minimum amount of fuel that must be stored onsite

is based on a new fuel consumption profile and ensures that

sufficient oil is present, even in the unlikely event that one EDG

storage tank (and its associated day tank) is unavailable. The

change to specification 3.7.F.4 is consistent with the newly

calculated amount of usable fuel and instrument uncertainties.

The deletion of the word ``available'' from Section 3.7.A.5

(concerning the individual storage tanks) and the change to

Reference 2 of Section 3.7 are administrative in nature and do not

involve a significant increase in the probability or consequences of

a previously analyzed accident.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

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

different kind of accident from any accident previously evaluated

because the changes do not affect current plant configuration or how

the plant operates. The proposed change in the minimum required

volume for the EDG fuel oil storage tanks ensures an adequate amount

of usable fuel and allows for temperature effects on fuel density

and calibration uncertainties. The change to the minimum amount of

fuel that must be stored onsite is based on a new fuel consumption

profile and ensures that sufficient oil is present, even in the

unlikely event that one EDG storage tank (and its associated day

tank) is unavailable. These changes do not alter how the fuel

storage tanks operate and therefore do not create the possibility of

a new or different kind of accident. Specification 3.7.F.4 is being

changed consistent with the revised calculation.

The deletion of the word ``available'' from Section 3.7.A.5

(concerning the individual storage tanks) and the change to

Reference 2 of Section 3.7 are administrative in nature and do not

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

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed changes do not involve a significant reduction in a

margin of safety. The proposed change in the minimum required volume

for the EDG fuel oil storage tanks ensures the required amount of

usable fuel is available for two EDGs to operate minimum safeguards

for 48 hours, and it allows for temperature effects on fuel density

and calibration uncertainties. The change to the minimum amount of

fuel that must be stored onsite is based on a new fuel consumption

profile and ensures that sufficient oil is present, even in the

unlikely event that one EDG storage tank (and its associated day

tank) is unavailable. Specification 3.7.F.4 is being changed

consistent with the revised calculation.

The deletion of the word ``available'' from Section 3.7.A.5

(concerning the individual storage tanks) and the change to

Reference 2 of Section 3.7 are administrative in nature and do not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Pao Tsin Kuo

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

James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: August 4, 1994

Description of amendment request: The proposed amendment would

modify the Technical Specifications (TSs) to revise the primary

containment atmosphere monitoring and drywell to torus differential

pressure requirements. Specifically, TS 3.7.A.6 would be revised to

adopt primary containment inerting/deinerting requirements that are

consistent with NUREG-1433, ``Standard Technical Specifications -

General Electric Plants, BWR/4.'' TSs 4.7.A.6.a and 4.7.A.7.a would be

revised to provide frequencies for the verifications of primary

containment oxygen concentration and pressure differential between the

drywell and torus. TSs 3.7.A.7.a.(1) and 3.7.A.7.a.(3) would be revised

to provide requirements for establishing and maintaining differential

pressure between the drywell and torus that are consistent with NUREG-

1433. Several administrative changes to Tables 3.2-8 and 4.3-8 were

also proposed to improve the overall quality of the TSs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

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

1. involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes revise primary containment atmosphere

monitoring requirements. The proposed changes adopt reference plant

operating conditions (i.e., 15% rated thermal power) for inerting/

de-inerting requirement as well as for the drywell to torus

differential pressure monitoring consistent with the NRC guidance

provided in the Standard Technical Specifications. The FitzPatrick

Technical Specifications currently allow a 24 [hour] grace period

following startup or before shutdown in which the primary

containment does not have to be inerted. During this 24 hour time

period required leak inspections as well as inerting or shutdown

evolutions are completed. Making the 24 hour ``window'' contingent

upon core thermal power will allow [operators] to place the mode

switch in run sooner, removing startup neutron monitoring

instrumentation scrams (i.e., APRM 15% and IRM upscale/inop). This

reduces the probability of spurious trips due to spiking of this

instrumentation. The proposed changes do not involve physical

modification to the plant nor involve any accident initiators.

Therefore, the probability of an accident occurring remains

unchanged. Accident analyses contained in FSAR [Final Safety

Analysis Report] Chapter 14 assume that a LOCA [Loss-of-coolant

accident] occurs from full power. The consequences of a LOCA below

15% rated thermal power would be less severe and would produce less

hydrogen.

The proposed changes to Tables 3.2-8 and 4.2-8 will eliminate

the reference to Specifications 3.7.A.9 by moving the primary

containment atmosphere monitoring requirements from Specification

3.7.A.9 to Table 3.2-8, Note F. Note F is also revised such that if

recorder 279CR-101A or B is inoperable, a daily monitoring and

logging of the appropriate parameter on the associated indicator on

panel 279CX-101A, B is acceptable in lieu of taking grab samples.

The monitoring will be performed using indicators on 279CX-101A and

B which are Regulatory Guide 1.97 qualified analyzers. The proposed

new Note K is added for completeness. These changes are

administrative in nature and will improve the overall quality of the

technical specifications. Therefore, the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes revise primary containment atmosphere

monitoring requirements by adopting STS [Standard Technical

Specifications] guidance regarding inerting/de-inerting

requirements. Consistent with this change the drywell to torus

differential pressure monitoring requirement is being revised.

Adopting the STS reference plant operating condition of 15% rated

thermal power adds operational flexibility. The proposed changes do

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

from any accident previously evaluated because the plant safety

analyses assume that a LOCA occurs at full power. In addition,

several changes are proposed to Tables 3.2-8 and 4.2-8 which

simplify hydrogen/oxygen monitoring requirements by moving the

primary containment monitoring requirements from Specification

3.7.A.9 to Table 3.2-8. These changes are administrative in nature

and will result in the overall improvement to the Technical

Specifications. Therefore, the proposed changes do not create the

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

previously evaluated.

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

The proposed changes revise the primary containment atmosphere

inerting/de-inerting requirements and the drywell to torus

differential pressure monitoring requirement. The proposed change

will allow inerting within 24 hours of exceeding 15% rated thermal

power during startup and de-inerting 24 hours prior to reducing

thermal power to less than 15% of rated before a plant shutdown.

These requirements are consistent with the guidance provided in the

STS. This proposed change does not affect the assumptions or

conclusions contained in the plant safety analyses which assume that

a LOCA occurs from full power. The consequences of a LOCA below 15%

rated thermal power would be less severe and would produce less

hydrogen. The proposed changes to Tables 3.2-8 and 4.3-8 are

administrative in nature. Therefore, the proposed changes do not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

York, New York 10019.

NRC Project Director: Pao Tsin Kuo

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 27, 1994

Description of amendment request: The amendment request proposes to

revise the Allowed Out-of-service Times (AOTs) for: inoperable Station

Service Water System (SSWS) pumps, inoperable Safety Auxiliaries

Cooling System (SACS) pumps, and inoperable Emergency Diesel Generators

(EDGs). In addition, this request is also proposing to allow online

maintenance of the EDGs.

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:

PSE&G has, pursuant to 10 CFR 50.92, reviewed the proposed

amendment to determine whether our request involves a significant

hazards consideration. We have determined that operation of the Hope

Creek Generating Station in accordance with the proposed changes:

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

consequences of an accident previously evaluated.

LCR 94-08

Station Service Water System (SSWS) Changes

Engineering evaluations of the SSWS/Safety Auxiliaries Cooling

System (SACS) demonstrate that adequate heat removal capability is

maintained in the post LOCA/LOP period with either two SSWS/SACS

pumps in one loop or with one SSWS/SACS pump in each independent

loop. The risk evaluations contained in the Probabilistic Safety

Assessment analyses of the SSWS determined that the probability of

an accident previously evaluated does not significantly change by

increasing the SSWS pump AOT from 7 days to 30 days. The evaluations

demonstrated that the relative risk remained low with an increased

(and more appropriate) AOT due to capabilities of the Hope Creek

SSWS to accommodate active failures.

Increasing the SSWS pump AOT does not involve physical

alteration of any plant equipment and does not affect analysis

assumptions regarding functioning of required equipment designed to

mitigate the consequences of accidents. Further, the severity of

postulated accidents and resulting radiological effluent releases

will not be affected by the increased AOT.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Safety Auxiliaries Cooling System Changes

Engineering evaluations of the SSWS/SACS demonstrate that

adequate heat removal capability is maintained in the post LOCA/LOP

period with either two SSWS/SACS pumps in one loop or with one SSWS/

SACS pump in each independent loop. The risk evaluations contained

in the Probabilistic Safety Assessment analysis of the SACS

determined that the probability of an accident previously evaluated

does not significantly change by increasing the SACS pump AOT from

72 hours to 30 days. Similarly, the provision of a 72 hour AOT for

one SACS pump inoperable in each SACS loop does not significantly

change the probability of an accident previously evaluated. The

evaluations demonstrated that the relative risk remained low with an

increased (and more appropriate) AOTs due to capabilities of the

Hope Creek SACS to accommodate active failures.

Increasing the SACS pump AOTs does not involve physical

alteration of the plant equipment and does not affect analysis

assumptions regarding functioning of required equipment designed to

mitigate the consequences of accidents. Further, the severity of

postulated accidents and resulting radiological effluent releases

will not be affected by the increased AOTs.

The proposed changes to ACTION Statement a.2 of Technical

Specification 3.7.1.1 precludes overly conservative and improper

operator action (initiation of plant shutdown procedures) to comply

with the requirements in the situation in which one of the affected

EDGs (an EDG cooled by the inoperable SACS loop) is not realigned to

OPERABLE SACS loop. Currently, Hope Creek can simultaneously be in

the ACTION Statement for Technical Specifications 3.7.1.1 and

3.8.1.1. Simultaneous entry into these ACTION Statements bounds the

conditions of the plant when the proposed requirements of the

Technical Specification 3.7.1.1, ACTION Statement a.2 are met. For

this reason, the proposed changes will not increase the

probabilities or consequences of an accident previously evaluated.

Technical Specification 3.7.1.1, ACTION Statements b., c. and d.

are being revised to require that the RHR loop or safety related

equipment must be declared inoperable when two SACS pumps in the

associated SACS loop are inoperable. This change permits one SACS

pump to be inoperable without affecting the operability of the

associated RHR loop or safety related equipment. Engineering

evaluations demonstrate that two SACS loops with one pump and two

heat exchangers per loop can provide the required heat removal

capability in the post DBA LOCA/LOP scenario and maintain safe

shutdown conditions. Therefore, a SACS loop with one OPERABLE SACS

pump should still be considered as a 100% functional SACS loop,

capable of supplying sufficient cooling for RHR and safety related

equipment required by Specifications 3.4.9.1, 3.4.9.2, 3.5.2,

3.9.11.1 and 3.9.11.2. For this reason, the proposed changes will

not increase the probabilities or consequences of an accident

previously evaluated.

In conclusion, the above SACS changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

LCR 94-11

Emergency Diesel Generator AOT Extensions.

The Hope Creek offsite and onsite power systems are highly

reliable. The risk evaluations contained in the Probabilistic Safety

Assessment analyses of the onsite power system determined that the

probability of an accident previously evaluated does not

significantly change by increasing the diesel generator AOT from 72

hours to 30 days for one inoperable diesel generator or from 2 hours

to 72 hours for two inoperable diesel generators. The evaluations

demonstrated that the relative risk remained low with an increased

(and more appropriate) AOT due to capabilities of the four channel

onsite Class-1E electrical system design at Hope Creek.

Increasing the diesel generator AOT does not involve physical

alteration of any plant equipment and does not affect analysis

assumptions regarding functioning of required equipment designed to

mitigate the consequences of accidents. Further, the severity of

postulated accidents and resulting radiological effluent release

will not be affected by the increased AOT.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

LCR 94-12

Emergency Diesel Generator Online Maintenance

The proposed changes would require that the requisite number of

diesel generators be in an operable condition, but would eliminate

the restriction that the 18 month maintenance inspection and other

surveillance tests be performed only while the unit is shutdown.

Because all operational conditions (governed by the operability of

the equipment prescribed as necessary in Technical Specification

3.8.1.1) and the associated actions are defined elsewhere in the

Technical Specifications, the removal of this restriction would not

involve as significant increase in the probability or consequences

of an accident previously evaluated.

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

accident from any accident previously evaluated.

LCR 94-08

Station Service Water System (SSWS) Changes

Extending the SSWS pump AOTS does not necessitate physical

alteration of the plant or changes in parameters governing normal

plant operation. Thus, this change does not create the possibility

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

evaluated for Hope Creek.

Safety Auxiliaries Cooling System Changes

The changes to the SACS do not necessitate physical alteration

of the plant or changes in parameters governing normal plant

operation. Thus, these changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated for Hope Creek.

LCR 94-11

Emergency Diesel Generator AOT Extensions

Extending the diesel generator AOTs does not necessitate

physical alteration of the plant or changes in parameters governing

normal plant operation. Thus, this change does not create the

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

previously evaluated for Hope Creek.

LCR 94-12

Emergency Diesel Generator Online Maintenance

The proposed revisions will not change the method in which any

of the 4.8.1.1.2.h surveillance activities are to be performed, only

the prescriptive operational condition is being removed. Since the

operational conditions and the associated actions are defined

elsewhere in the Technical Specifications, the removal of this

restriction will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

safety.

LCR 94-08

Station Service Water System (SSWS) Changes

As discussed above, the Probabilistic Safety Assessment analyses

determined that the change in core damage frequency for extended

SSWS pump AOT is insignificant. Therefore, this change does not

result in a significant reduction in a margin of safety.

Safety Auxiliaries Cooling System Changes

As discussed above, the Probabilistic Safety Assessment analyses

determined that the change in core damage frequency for the SACS

changes are insignificant. Therefore, these changes do not result in

a significant reduction in a margin of safety.

LCR 94-11

Emergency Diesel Generator AOT Extensions

As discussed above, the Probabilistic Safety Assessment analyses

determined that the change in core damage frequency for extended

diesel generator AOTs is insignificant. Therefore, this change does

not result in a significant reduction in a margin of safety.

LCR 94-12

Emergency Diesel Generator Online Maintenance

The margin of safety for the emergency power system depends on

the proven, historical reliability of the diesel generators and the

surveillances verifying the power circuits between the offsite and

the onsite power systems. The elimination of the restrictions for

performance of the maintenance tear down inspection would remain

within the action parameters of Technical Specification 3.8.1.1.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: Mohan C. Thadani, Acting

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 28, 1994

Description of amendment request: The proposed Technical

Specification changes contained herein represent changes to Section 3/

4.8.1 ``AC Sources.'' The revised specification removes the

surveillance requirements, methodology and frequency for Emergency

Diesel Generator (EDG) fuel oil from the Technical Specifications and

relocates them in a controlled plant procedure, VSH.SS-CA.ZZ-0013(Q)

``Procedure for Testing Diesel Fuel and 2 Fuel Oil at

Artificial Island for PSE&G Nuclear Operations.'' The changes also

delete an unnecessary lab test for the fuel oil and extend the

surveillance frequency from once per 92 days to once per 184 days.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

The proposed change to delete a test not required by Regulatory

Guide 1.137 or ASTM-D975-77 will not result in degradation of fuel

oil quality below acceptable limits. Based on established fuel oil

quality history, the proposed increase in surveillance frequency

from once per 92 days to once per 184 days will not significantly

decrease confidence in fuel oil quality and EDG operability, nor

will the relocation of fuel oil quality surveillance from the

Technical Specifications to the Diesel Fuel Oil Testing Program have

any effect on established plant practices in regards to the testing

of EDG fuel oil. The proposed changes involve no hardware changes,

no changes to the operation of any systems or components, and no

changes to existing structures. Therefore, these changes will not

alter or impact previously evaluated accidents.

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

accident from any previously evaluated.

The proposed changes are procedural in nature concerning fuel

oil testing and, therefore, will not directly impact the operation

of any plant safety related component or equipment. Any reduction in

fuel oil quality will not be significant or result in a decrease in

EDG operability. Therefore, these changes will not create a new or

unevaluated operating condition.

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

safety.

The proposed changes concern how EDG fuel oil quality is to be

determined, how frequently this determination is to be performed,

and how to control the process for determining fuel oil

acceptability, and therefore EDG operability. There are no

associated safety margins and the only margin of concern is that of

fuel oil combustibility due to the presence of either contaminants

or particulate buildup from long term storage. Based on historical

data, PSE&G believes that EDG fuel oil quality will not be affected

or impacted by the proposed changes. Therefore, the proposed

amendment does not involve any reduction in a safety margin.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: Mohan C. Thadani, Acting

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

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

San Diego County, California

Date of amendment requests: May 20, 1994.

Description of amendment requests: This is a proposal to revise the

Units 2 and 3 Technical Specification (TS) 3/4.7.3, ``Component Cooling

Water System,'' and the corresponding Bases to support the addition of

the component cooling water surge tank backup nitrogen supply (BNS)

system. The amendment is necessary to establish new operability and

surveillance requirements for the system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The Component Cooling Water (CCW) system removes heat to

mitigate the consequences of those design basis accidents included

in chapter 15 of the Updated Final Safety Analysis Report (UFSAR). A

CCW system failure is not an accident initiating event as listed in

the UFSAR, Table 15.0-2. The addition of the Backup Nitrogen Supply

(BNS) system does not change the CCW system function and does not

interface with any system which relates to the initiating events

listed in Table 15.0-2 of the UFSAR. The BNS system is designed to

Quality Class II, Seismic Category I requirements and will increase

CCW reliability by minimizing CCW system voiding during and after a

Design Basis Event (DBE). Failure of the BNS system will not by

itself result in an accident or have any effect on normal plant

operation.

The proposed revision of Technical Specification (TS) 3/4.7.3

will not change the CCW system operation. This amendment request

retains the original CCW TS requirements and adds provisions

specifically limited to the BNS system. The proposed revisions

provide an 8-hour Allowed Outage Time (AOT) for one or both trains

of the BNS system inoperable to avoid unnecessary plant power

reductions. If the 8-hour AOT for BNS system inoperability is not

met, the associated CCW train(s) must be declared inoperable. The 8-

hour AOT followed by either the 72-hour AOT for one train of CCW

inoperable or the 1-hour AOT provided by TS 3.0.3 for both trains of

CCW inoperable results in overall AOTs of 80 and 9 hours,

respectively. The results of a conservative Probabilistic Risk

Assessment demonstrate that for the overall 80-hour and 9-hour AOTs

the increases in core damage risk per year are 6.5E-7 and 8.6E-7,

respectively. This results in less than a 3% increase in the annual

core damage risk for Units 2 and 3.

The proposed revisions to TS 3/4.7.3 include surveillance

requirements to provide assurance that the BNS system remains

OPERABLE when required to support CCW operation. Therefore,

operation of the facility in accordance with this proposed TS change

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this

proposed amendment create the possibility of a new or different kind

of accident from any accident previously evaluated?

Response: No

The BNS system does not change the CCW system function and does

not interface with any system related to the initiating events

listed in Table 15.0-2 of the UFSAR. The BNS system is designed to

Seismic Category I requirements and will minimize CCW system voiding

and the potential for a subsequent water hammer by maintaining the

CCW surge tank pressure during and after a DBE. No new High Energy

Line Break considerations apply because the nitrogen bottle pressure

is reduced at the bottle header and all connections are less than

one inch in diameter. The BNS system is independent from all systems

possibly related to the initiating DBEs listed in the UFSAR Table

15.0-2.

The proposed TS 3/4.7.3 revision does not change the existing

CCW system requirements. This proposed change adds operability and

surveillance requirements for the BNS system to support CCW system

operability and provide additional assurance that plant operation is

consistent with the design basis. Failure of the BNS system will not

by itself result in an accident or have any effect on normal plant

operation. Therefore, the proposed changes will not create the

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

previously evaluated.

3. Will operation of the facility in accordance with this

proposed amendment involve a significant reduction in a margin of

safety?

Response: No

The addition of the BNS system enhanced the CCW system by

minimizing the possibility for water hammer following certain

postulated events. Surveillance and nitrogen bottle change-out

procedures assure that the BNS system is available to perform its

safety-related function. The redundant cooling capacity of the CCW

system is maintained by providing an independent dedicated BNS

system for each CCW critical loop, assuming a single failure.

The safety function of the BNS system is limited to the

minimization of void formation in the CCW system under a specific

set of coincident circumstances following a DBE. The proposed

revision to TS 3/4.7.3 allows the BNS system to have one or both

trains inoperable for 8 hours before the associated CCW train(s)

must be declared inoperable. The BNS system AOTs do not affect plant

operation because the BNS system is not normally in operation. The

BNS system action statements are not normally entered for normal

bottle change out since the BNS system is designed with one more

bottle than is required for seven days of BNS system operation.

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

margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

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

Attorney for licensee: James A. Beoletto, Esquire, Southern

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

NRC Project Director: Theodore R. Quay

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

Callaway County, Missouri

Date of amendment request: May 20, 1994

Description of amendment request: The proposed amendment would

modify the Technical Specifications to incorporate improvements

endorsed by the NRC Final Policy Statement on Technical Specification

Improvements for Nuclear Power Reactors.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed Technical Specification changes involve relocating

requirements that are not conditions or limitations on reactor

operation necessary to obviate the possibility of an abnormal

situation or event giving rise to an immediate threat to the public

health and safety. The proposed changes were identified through the

application of criteria designed to cull those requirements that are

not important to operational safety from the Technical

Specifications. In this process, selected provisions of the

Technical Specifications identified for relocation were retained if

necessary to support a Technical Specification that was to be

retained. Thus, only specification requirements that have little or

no operational safety significance are proposed for relocation. In

addition, those requirements that would be relocated will be

included in the Final Safety Analysis Report (FSAR) and, therefore,

will be controlled and implemented as FSAR commitments. In this

manner, those requirements that have no operational safety

significance but involve maintaining the plant in its as-designed

state, (for example, through surveillance programs) would be

controlled.

In addition, the criteria for identifying requirements to be

retained in Technical Specifications specifically call out, for

retention, those structures, systems, or components that are

required to mitigate accidents previously evaluated.

Based on the above, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any previously evaluated.

The proposed changes involve relocating Technical Specification

requirements to another licensee-controlled document, i.e. FSAR

Chapter 16. No changes or physical alterations of the plant are

involved. Also, no changes to the operation of the plant or

equipment are involved. Therefore, the proposed changes do not

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

any previously evaluated.

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

The proposed changes involve relocating Technical Specification

requirements to the FSAR. The requirements to be relocated were

identified by applying the criteria endorsed in the Commission's

Policy Statement. Thus, those specifications that would be relocated

do not impose constraints on design and operation of the plant that

are derived from the plant safety analysis report or from

probabilistic safety assessment (PSA) information and do not belong

in the Technical Specifications in accordance with 10 CFR 50.36 and

the purpose of the Technical Specifications stated in the Policy

Statement. Therefore, relocation of these requirements does not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: John N. Hannon

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

Callaway County, Missouri

Date of amendment request: June 21, 1994

Description of amendment request: The proposed amendment would

delete the sections describing the On-Site Review Committee (ORC) and

the Nuclear Safety Review Board (NSRB) from the Technical

Specifications. This change also removes reference to the Manager,

Nuclear Safety and Emergency Preparedness. Additionally, the change

reflects an organizational restructuring which addresses the

Independent Safety Engineering Group (ISEG) reporting to the Manager,

Quality Assurance.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The changes are administrative and equivalent descriptions and

requirements for these oversight committees are contained in FSAR

Section 13.4.

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

from any previously evaluated.

These changes do not involve any physical alterations to the

plant. There is no new type of accident or malfunction created and

the method and manner of plant operation will not change. The

changes are administrative and equivalent descriptions and

requirements for these oversight committees are contained in FSAR

Section 13.4.

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

The margin of safety remains unaffected since no design change

is made and plant operation remains the same. The changes are

administrative and equivalent descriptions and requirements for

these oversight committees are contained in FSAR Section 13.4.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: John N. Hannon

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

Project No. 2, Benton County, Washington

Date of amendment request: July 12, 1994

Description of amendment request: The proposed amendment would

modify the technical specifications (TS) to remove instrument response

time limit tables for the reactor protection system (RPS) and isolation

actuation and emergency core cooling system (ECCS) from the TS. The

affected instrument response time limit tables would be located in the

Final Safety Analysis Report (FSAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The RPS, Isolation Actuation and ECCS Instruments provide

signals to the actuation logic for safety equipment needed to

mitigate accidents and transients. The proposed change relocates the

instrument response times from the Technical Specifications to the

FSAR but will not affect the operability or surveillance

requirements of the affected instruments. The instruments will

continue to be proven operable on the schedule provided in the

Technical Specifications.

The FSAR change process and Plant Operations Committee review

responsibilities ensure that changes to the response time limits

cannot be made without adequate review and approval. Since

operability confirmation as required by the Technical Specifications

(surveillance testing requirements) will not be affected by the

change and the limits themselves cannot be altered without adequate

review and approval, there is no possibility of a significant

increase in the probability of an accident previously approved as a

result of this change.

The instruments provide signals to the actuation logic of

equipment used to mitigate the consequences of an accident. However,

since no changes are being made in the methods or frequencies of

proving operability the systems will not be degraded or be made

susceptible to degradation that could go unidentified. As discussed

above, changes to the limits will not be made without adequate

review and approval. Hence, this change will not affect the

capability of the plant to mitigate a previously evaluated accident.

Because the mitigative capability is not affected there is no

significant increase in the consequences of a previously evaluated

accident as a result of this change.

For the above reasons, the change does not represent a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change relocates only the tables containing the

instrument response times for the RPS, Isolation Actuation and ECCS

response time limits from the Technical Specifications to the FSAR.

The change does not affect how these instruments will function.

Relocation of this information does not represent a change in the

configuration or operation of the plant. No new hardware is being

added to the plant as part of the proposed change. Plant procedures

are not affected by the change. The Technical Specification sections

for the surveillance testing of these instruments will not be

affected. Therefore, the Technical Specifications will continue to

require that the same operability and surveillance requirements be

met for the affected instruments.

Consequently, the possibility of a new or different kind of

accident from any accident previously analyzed is not introduced as

a result of this change.

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

of safety?

The margin of safety established by the response time limits is

in ensuring that the RPS, Isolation Actuation and ECCS systems will

respond in time to support the assumptions of the accident analysis.

Relocating the response time limits to the FSAR does not alter the

operability or the surveillance requirements applicable to the

affected instruments. These instruments will continue to be tested

for operability and therefore remain capable of responding to

accident events within the time limits required by the accident

analysis. The administrative change control provisions for the FSAR,

the plant procedures implementing the requirements of 10 CFR 50.59

and the administrative sections of the Technical Specifications are

adequate to control changes to the response time limits such that

they cannot be altered in a manner that would adversely affect plant

safety.

Therefore, for these reasons, the change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005-3502

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.

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

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

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of amendment request: March 23, 1994, as supplemented on July

26, 1994

Description of amendment request: The proposed amendments consist

of two parts: Part one, would revise ``Moderator Temperature

Coefficient (MTC)'' Technical Specifications (TSs) to allow the use of

a slightly positive MTC for the core design. The licensee has stated

that a positive MTC will reduce the burnable rod requirements and

improve operational flexibility. Because of using a positive MTC, the

TSs would be revised to permit a higher boron concentration in the

refueling water storage tank, the reactor coolant system (RCS)

accumulators, and the refueling cavity, in order to ensure adequate

shutdown margin is maintained at all times. Part two, would revise the

TSs to reduce the required RCS flow to offset any reduction in flow due

to increased steam generator tube plugging. Additionally, the

associated Bases for the above TSs would be revised to describe the

basis for the TS requirements.

Because Byron, Unit 1, and Braidwood, Unit 2, will be in refueling

outage in the fall of 1994, the proposed TS changes will apply to them.

Byron, Unit 2 and Braidwood, Unit 1 will continue to operate in

accordance with the current TSs. The licensee's submittal identified

the appropriate unit applicability of the TSs pertaining to the

positive MTC and the required RCS flows.Date of publication of

individual notice in Federal Register: August 15, 1994 (59 FR 41802)

Expiration date of individual notice: September 14, 1994

Local Public Document Room location: For Byron, the Byron Public

Library, 109 N. Franklin, Byron, Illinois 61010; and for Braidwood, the

Wilmington Township Public Library, 201 S. Kankakee Street, Wilmington,

Illinois 60481.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

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

Illinois

Date of amendment request: August 5, 1994

Brief description of amendment request: The proposed amendment

would modify Technical Specification Table 4.8.1.1.2-1, ``Diesel

Generator Test Schedule,'' to exclude selected valid failures of the

Division 1 diesel generator from contributing to an accelerated testing

frequency.Date of publication of individual notice in Federal Register:

August 16, 1994 (59 FR 42080).

Expiration date of individual notice: September 15, 1994

Local Public Document Room location: Vespasian Warner Public

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

Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and

Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606

NRC Project Director: John N. Hannon

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

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

Pennsylvania

Date of amendment request: January 14, 1994

Brief description of amendment request: The proposed amendment

would increase the storage capacity in each spent fuel pool from their

current 2040 fuel assemblies to 4117 fuel assemblies. In addition, the

proposed amendment would extend the ``fuel core reserve'' capability

from year 1998 to 2013.

Date of publication of individual notice in Federal Register:

August 8, 1994 (59 FR 40376)

Expiration date of individual notice: September 7, 1994

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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 Commi

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