U.S. NUCLEAR REGULATORY COMMISSION

NATIONAL FOUNDATION ON THE ARTS AND THE HUMANITIES

U.S. NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 16, 1994, through May 26, 1994. The last

biweekly notice was published on May 25, 1994 (59 FR 27049).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

delivered to Room 6D22, Two White Flint North, 11555 Rockville Pike,

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

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

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

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

intervene is discussed below.

By July 8, 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.

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

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

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: December 30, 1993

Description of amendment requests: The proposed amendment would

modify Tables 3.3-4 and 3.3-5 of Technical Specification 3/4.3.2,

``Engineered Safety Features Actuation System Instrumentation,'' to

provide clarification of settings for undervoltage relay trip values

for the Class IE 4.16 kV electrical bus. The proposed amendment would

also add Figure 3.3-1, ``LOSS OF VOLTAGE RELAY (GE IAV) TIME VS VOLTAGE

CURVE,'' to clarify the relay setpoints and methodology.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1--Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed TS amendment does not significantly increase the

probability of an accident previously evaluated. The methodology

remains the same.

Clarifying the minimum acceptable voltage and allowing more

conservative values of the 4.16 kV bus undervoltage trip value will

ensure that 4.16 kV ESF [Engineered Safety Features] bus voltages

are sufficient to provide adequate voltage to equipment necessary

for accident response.

Standard 2--Create the possibility of a new or different kind of

accident from any accident previously analyzed.

The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously analyzed.

No new or different methodology is being proposed.

The minimum undervoltage relay setpoints are specified and the

new wording will allow a more conservative setpoint to ensure

voltage levels to equipment powered by the 4.16 kV ESF bus.

The proposed amendment also adds a figure to clarify the relay

setpoints and methodology.

Standard 3--Involve a significant reduction in a margin of

safety.

The margin of safety as defined in the TS will be increased by

specifying the minimum and allowing more conservative values for the

4.16 kV ESF bus undervoltage relay trip values. This clarification

will ensure that the trip setpoint adequately protects that

equipment powered by the 4.16 kV ESF bus from a potentially damaging

degraded voltage condition.

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

that 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: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

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

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

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: March 28, 1994

Description of amendment requests: The proposed amendment would

change Technical Specification 3/4.3.7.1.3 Condensate Storage Tank. The

licensee proposed to change the minimum condensate storage tank (CST)

indicated level from 25 feet to 29.5 feet to ensure that the CST

contains sufficient volume of water. In addition, the licensee proposed

to make an editorial change to the Unit 3 Technical Specification 3/

4.3.7.1.3, from ``with a level'' to ``with an indicated level,'' to be

consistent with the Units 1 and 2 Technical Specifications.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1--Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This amendment request does not involve a significant increase

in the probability or consequences of an accident previously

evaluated based on the safety analysis for the CST [condensate

storage tank] minimum indicated level. The proposed change increases

the minimum indicated CST water level from 25 feet to 29.5 feet.

Increasing the minimum indicated CST water level ensures that the

requirements of BTP [Branch Technical Position] RSB [Reactor System

Branch] 5-1 and UFSAR [Updated Final Safety Analysis Report] Section

9.2.6 continue to be met. Therefore, this proposed change ensures

that the consequences of an accident previously evaluated are not

affected.

Standard 2--Create the possibility of a new or different kind of

accident from any accident previously analyzed.

This amendment request does not create the possibility of a new

or different kind of accident from any accident previously analyzed

since the minimum water volume of 300,000 gallons is maintained by

this change. The change in the CST minimum required water level does

not change the operation of any plant equipment while ensuring that

the required 300,000 gallons are available. Since this change does

not affect the operation of plant equipment and ensures that the

minimum CST water inventory is maintained, the proposed change does

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

from any accident previously analyzed.

Standard 3--Involve a significant reduction in a margin of

safety.

The margin of safety presently provided is not reduced by the

proposed change in the CST minimum required water level. The

proposed change ensures that the CST volume of 300,000 gallons is

available to satisfy the requirements of BTP RSB 5-1, UFSAR Section

9.2.6 and the BASES for Technical Specification 3/4.3.7.1.3.

Therefore, since the minimum required CST level in maintained, the

proposed change does not involve a significant reduction in a margin

of safety.

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

that 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: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

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: May 20, 1994

Description of amendment request: The proposed amendment will

change the Technical Specifications (TS) to (1) add an exception to TS

6.3.1 regarding the requirement for the Manager - Operations position

to hold a Senior Reactor Operator's (SRO's) license, (2) add a new TS

6.3.2 to describe the qualifications for the Manager - Operations and

Manager - Shift Operations positions. This new section will require the

Manager-Operations to hold or have held an SRO license for either the

H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson), or a

similar plant and require the Manager-Shift Operations to hold an SRO

license at the Robinson plant, and (3) renumber the sections to allow

for the insertion of a new TS 6.3.2.

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:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The total number of senior reactor operator

licensed personnel on shift remains unchanged. The change to provide

a middle level of management will lessen the burden of the daily

shift operations placed on the Manager - Operations. This will allow

for more effective overall management of the Operations Unit.

Requiring the Manager shift Operations to hold an SRO license will

assure that supervision of operator activities continues to be

performed by a senior licensed individual. The proposed changes to

the Operations management organization do not involve physical

alterations of the plant configuration or changes in setpoints or

operating parameters. 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. As described above, these changes affect the organization

of the Operations Unit. They do not represent any appreciable change

in the current methodologies; they merely update the TS to reflect

current personnel organization configuration and standards. The

proposed changes to the Operations management organization do not

involve physical alterations of the plant configuration or changes

in setpoints or operating parameters. Therefore, the changes

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

accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety. The changes proposed do not

reduce the number of senior reactor operator licensed personnel on

shift. The changes to the Operations organization, as reflected in

the proposed change, will enhance the overall effectiveness of plant

operations and will serve to improve nuclear safety. There are no

changes to the plant configuration or changes in setpoints or

procedures. Therefore, the changes proposed have no affect on the

facility's 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,

Home and Fifth Avenues, 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: William H. Bateman

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

Michigan

Date of amendment request: May 10, 1994

Description of amendment request: The proposed amendment would

remove component lists from Technical Specification (TS) sections in

accordance with the guidance provided in Generic Letter (GL) 91-08

dated May 6, 1991. Related TS which reference the lists are also

modified.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed change will not result in any hardware or operating

changes. The proposed change is based upon Generic Letter 91-08 and

merely removes component lists, removes details relating to the

component lists, provides clarifying information supporting the

removal of the component listings, or removes details (which are

considered administrative) that are no longer applicable to the TS.

The removal of tabular component listings from the TS does not

impact affected component OPERABILITY requirements. TS will continue

to require the components to be OPERABLE. Action statements and

surveillance requirements for the components will also remain in the

TS. The tabular component lists will be relocated to plant

procedures which will be controlled in accordance with the

provisions specified in the Administrative Controls Section of the

TS. Therefore, this change is administrative in nature and does not

involve a significant increase in the probability or consequences of

an accident previously evaluated. Further, the proposed changes do

not alter the design, function, or operation of the components

involved and therefore, do not affect the consequences of any

previously evaluated accident.

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

or different kind of accident from any accident previously

evaluated. The proposed changes will not impose any different

operational or surveillance requirements. The changes propose to

relocate these component lists to plant procedures whereby adequate

control of information is maintained. Further, as stated above, the

proposed changes do not alter the design, function, or operation of

the components involved and therefore, no new accident scenarios are

created.

3. The proposed changes do not involve a significant reduction

in a margin of safety. The proposed change will not reduce a margin

of safety because it has no impact on any safety analysis

assumption. The proposed change does not alter the scope of

equipment currently required to be OPERABLE or subject to

surveillance testing nor does the proposed change affect any

instrument setpoints or equipment safety functions. Therefore, 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

2000 Second Avenue, Detroit, Michigan 48226

NRC Project Director: Ledyard B. Marsh

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

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

Date of amendment request: July 15, 1993

Description of amendment request: The revision proposed by

Technical Specification Change Request (TSCR) No. 211 to the Technical

Specifications would remove a footnote to specification 3.24, ``Reactor

Coolant Inventory Trending System,'' and would revise that

specification to be consistent with the revised Babcock & Wilcox

revised Standard Technical Specifications, issued as NUREG-1430.

The July 15, 1993, request supersedes the request of September 5,

1991, published in the Federal Register on November 13, 1991 (56 FR

57697).

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 revised Limiting Condition for Operation represents no

increase in the probability of occurrence or consequences of an

accident previously evaluated.

The TSCR represents no change to the physical configuration or

operation of the Reactor Coolant Inventory Trending System.

As stated in the Bases for the existing Technical Specification

3.24, the system is not a system required to mitigate accidents. It

may be useful to have the system operable, but adverse impact does

not result if it is not operable; other useful information for

monitoring inadequate core cooling is available. The change proposed

is in accordance with the Limiting Condition for Operation in NUREG

1430.

2. The revised Limiting Condition for Operation does not create

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

accident previously evaluated.

As identified above, the TSCR represents no change to the

physical configuration or operation of the Reactor Coolant Inventory

Trending System.

3. The revised Limiting Condition for Operation does not involve

a significant reduction in a margin of safety.

The margin of safety for the proposed Limiting Condition for

Operation is no different from that for existing Technical

Specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz

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

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

Date of amendment request: April 11, 1994

Description of amendment request: The purpose of the request is to

revise the Technical Specifications (TS) by relocating the detailed

inspection criteria, methods and frequencies of the containment tendon

surveillance program to the Final Safety Analysis Report (FSAR) and

providing a direct reference to the existing tendon surveillance

program in the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed amendment only relocates the

tendon surveillance program detailed requirements and criteria to

the FSAR, consistent with the BWOG [Babcock & Wilcox Owners Group]

Revised Standard Technical Specifications. The proposed amendment

does not affect the requirement to verify the containment structural

integrity in accordance with the inservice tendon surveillance

program. The proposed Technical Specification specifies that the

tendon surveillance program conforms to the recommendations of U.S.

NRC Regulatory Guide 1.35, which ensures the effectiveness of the

program is not reduced and containment structural integrity is not

affected. Therefore, this change does not increase the probability

of occurrence or the consequences of an accident previously

evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated. The tendon

surveillance program is required to conform to the recommendations

of U.S. NRC Regulatory Guide 1.35. Therefore, the effectiveness of

the surveillance program is maintained, thus providing continued

assurance of containment structural integrity.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The performance of the inservice tendon surveillance program

is maintained in conformance with the recommendations of U.S. NRC

Regulatory Guide 1.35, thus providing continued assurance of

containment structural integrity. Therefore, it is concluded that

operation of the facility in accordance with the proposed amendment

does not involve a reduction in a margin of safety as defined in the

basis of any Technical Specification.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz

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

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

Date of amendment request: April 19, 1994

Description of amendment request: The purpose of the request is to

delete the quality assurance (QA) audit program frequency requirements

from the Technical Specifications (TS) and to utilize the Operational

Quality Assurance (OQA) Plan as the controlling document.

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. These changes do not affect the function of any system or

component. Therefore, they do not increase the probability of

occurrence or the consequences of an accident previously evaluated

in the SAR [Safety Analysis Report].

2. These changes do not involve a physical change to plant

configuration and they do not affect the performance of any

equipment. Therefore, they do not create the possibility of a new or

different kind of accident or malfunction of a different type than

previously identified.

3. The shifting of the audit frequency requirements from the

Technical Specifications to the OQA Plan and the extension of the

maximum interval between audits of certain areas do not change the

activities to be audited nor the scope of individual audits.

Furthermore, audit frequencies are not associated with the marginof

safety in the bases of any Technical Specification. Therefore, the

margin of safety in not affected by this change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz

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: April 14, 1994.

Description of amendment request: The licensee proposes to revise

Technical Specification 5.3.1, ``Fuel Assemblies.'' The amendment would

permit fuel assembly reconstitution to restore the usefulness of fuel

assemblies containing damaged or leaking fuel rods.

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.

Fuel assemblies containing filler rods will be shown to meet the

current nuclear, mechanical, and thermal-hydraulic design limits on

a cycle-specific basis. Since fuel assemblies containing filler rods

will be shown to meet the current nuclear, mechanical, and thermal-

hydraulic design limits on a cycle-specific basis, there is no

impact on the design basis of the plant.

Replacement of fuel rods with fuel rods containing natural

uranium or fuel rods from fresh or burned assemblies will be

evaluated by South Texas Project's internal 10 CFR 50.59 review

process. Reconstituted fuel assemblies with fuel rods containing

natural uranium and/or fuel rods from fresh or burned assemblies

will be analyzed using the normal design methodology as described in

Section 6.9.1.6 of the Technical Specifications.

Since current reload core design limits will be met by fuel

assemblies using zirconium alloy or stainless steel filler rods, the

proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind accident from any previously evaluated.

The use of filler rods does not involve any alteration to plant

equipment or procedures which would introduce any new or unique

operational modes or accident precursors. Therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The use of filler rods does not change the reload design or

safety analysis limits for a reload core. Their use will be

evaluated on a cycle-specific basis using NRC-accepted fuel rod

configurations and analysis techniques. Since the safety analysis

limits are unaffected and since the modified fuel assemblies will be

shown to meet existing design limits on nuclear, mechanical, and

thermal-hydraulic parameters, 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 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: James E. Lyons, 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: April 29, 1994

Description of amendment request: The licensee proposes to revise

Technical Specification 3.7.1.1, ``Turbine Cycle - Safety Valves.'' The

amendment would change the maximum allowable power range neutron flux

high setpoint when one or more main steam safety valve (MSSV) is

inoperative. The Bases for Technical Specification 3.7.1.1 would also

be changed to reflect the new algorithm used to calculate the revised

setpoint values.

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 Main Steam Safety Valves [MSSVs] are only actuated

after a transient has occurred. Therefore, revising the maximum

power level with inoperable MSSVs would not increase the probability

of a previously evaluated accident. Following a loss of load/turbine

trip, the revised Technical Specification Table 3.7-1 maximum

allowable power range neutron flux high setpoints would ensure that

the maximum power level allowed for operation with inoperable MSSVs

is below the heat removing capability of the operable MSSVs. This

would ensure that the design limit of 110% overpressurization is not

exceeded. Therefore, there is no increase in the consequences of an

accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed change would not alter the design, configuration,

or method of operation of STP [South Texas Project]. For this

reason, as well as the reasons stated in response to Criterion 1

above, the proposed change would not create the possibility of a new

or different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The revised Table 3.7-1 setpoint values were calculated using a

conservative method where the maximum power level allowed for

operation with inoperable MSSVs is below the heat removing

capability of the operable MSSVs. Using the revised maximum plant

operating power levels will ensure that the secondary system

pressure will be limited to within 110% of its design pressure.

Therefore, since the design criteria will continue to be met, 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 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: Suzanne C. Black

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

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

Pennsylvania

Date of amendment request: April 27, 1994

Description of amendment request: The amendment would delete

Technical Specifications Surveillance Requirement 4.4.1.1.1, which

requires that the reactor recirculation pump discharge valve be

demonstrated operable by performing a full-stroke test of the valve

prior to reactor thermal power exceeding 25% of rated reactor thermal

power.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed change to delete the TS Surveillance Requirement

(SR) does not require any modifications to the plant or equipment,

and will not impact the operation of the reactor recirculation

system. The reactor recirculation system will continue to function

as designed to maintain reactor pressure boundary integrity and to

provide sufficient flow through the reactor core to remove heat from

the fuel. This proposed TS change does not affect the operation of

any Emergency Core Cooling System (ECCS) or other plant equipment

important to safety. The purpose of this TS SR is to satisfy an ECCS

operability requirement for Boiling Water Reactor (BWRs) where the

reactor recirculation system piping serves as the injection flowpath

to the reactor pressure vessel for the Low Pressure Coolant

Injection (LPCI) system, an ECCS. Each LPCI subsystem, at LGS, has

an independent flowpath which does not rely on the reactor

recirculation system piping for injecting to the reactor pressure

vessel. The reactor recirculation system pump discharge valves do

not perform an active safety-related function and are classified as

passive safety-related components designed to maintain the reactor

pressure boundary integrity during reactor recirculation pump

maintenance activities. These valves are not normally used during

plant operations except to establish normal shutdown cooling, a

manually initiated non-safety related function. These valves are not

used to mitigate the consequences of design bases accidents.

Therefore, the proposed change does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

This proposed TS does not require physical changes to the plant

or equipment, and does not impact any design or functional

requirements of the reactor recirculation system, LPCI system, or

other plant systems important to safety. The purpose for this TS SR

is to satisfy an ECCS operability requirement for BWRs where the

reactor recirculation system piping serves as the injection flowpath

to the reactor pressure vessel for the LPCI system. The LPCI system

at LGS has an independent flowpath for injecting to reactor pressure

vessel and does not rely on the reactor recirculation system piping

as part of the injection flowpath. Since the intent of this

requirement is to support LPCI operation, and the LPCI system design

function is accident mitigation, eliminating this TS SR has no

impact on the types of accidents that could occur. The reactor

recirculation system pump discharge valves do not perform an active

safety-related function and are classified as passive safety-related

components designed to maintain the reactor pressure boundary

integrity during reactor recirculation pump maintenance activities.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The proposed change to delete the TS SR does not involve a

change to the physical design or functional requirements of the

reactor recirculation system, LPCI system, or other plant system

important to safety. The reactor recirculation system will continue

to function as designed to maintain reactor pressure boundary

integrity and to provide sufficient flow through the reactor core to

remove heat from the fuel. This proposed TS change does not impact

the safety-related operation of the LPCI system. The LPCI system

will continue to function as designed to mitigate the consequences

of an accident. These valves do not perform an active safety-related

function and are classified as passive safety-related components

designed to maintain the reactor pressure boundary integrity during

reactor recirculation pump maintenance activities.

Therefore, the proposed TS change does not involve a reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller

Philadelphia Electric Company, Public Service Electric and Gas

Company,Delmarva Power and Light Company, and Atlantic City

Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: April 27, 1994

Description of amendment request: These technical specifications

(TS) changes are being proposed to support the implementation of

proposed Modification 5274 which is intended to replace the Peach

Bottom Atomic Power Station (PBAPS), Unit 2 Containment Atmospheric

Dilution (CAD) System and Containment Atmospheric Control (CAC) System

analyzers. This modification was performed on PBAPS, Unit 3 during the

previous Unit 3 outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The design function and operation of the CAC and CAD Systems,

which are supported by the operation of the containment monitoring

system, have not been altered as a result of these changes. The CAC

System monitors the content of oxygen during startup and normal

operation and the CAD System is utilized to monitor the content of

hydrogen and oxygen during post-LOCA [loss-of-coolant accident]

operation. The monitoring of these variables will continue to

mitigate the consequences of accidents previously evaluated.

Additionally, no accident precursors will be impacted by these

changes.

The new system meets or exceeds the design standards of the

original system. Additionally, the decrease in warmup time will

increase the availability and usefulness of the analyzers to

mitigate the consequences of an accident. Therefore, the proposed

changes will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

from any accident previously evaluated; or,

The proposed TS changes do not involve the introduction of any

new accident initiators. The new containment monitoring system will

enhance the ability of [the] CAD system to mitigate the consequences

of an accident and prevent the introduction of a new or different

type of accident previously evaluated. The new system meets or

exceeds existing design standards and will be tested to ensure its

reliability. The new containment monitoring system is a monitoring

system and will not introduce new accident initiators. Therefore,

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

different type of accident from any accident previously evaluated.

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

Although the number of analyzers is being reduced, the proposed

modification and TS changes will enhance the ability of the

containment monitoring system to support the operation of the CAC

and CAD systems [through] the use of improved equipment that meets

or exceeds the design standards of the original system. Therefore,

the proposed changes will not reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller

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: May 10, 1994

Description of amendment request: The proposed amendment would

revise Section 3.1.C.3 and Table 4.1-1 of Appendix A of the Operating

License. This change would require that the reactor coolant average

temperature (Tavg) be restored to greater than or equal to

540 deg. F within a 15-minute period or be in hot shutdown within the

following 15 minutes. The proposed change in Table 4.1-1 entitled,

``Minimum Frequencies for Checks, Calibrations and Tests,'' will add

the requirement for Tavg instrument check frequency to be reduced

to 30 minutes when the Tavg deviation and low Tavg alarms are

not reset and the control banks are above zero steps. This application

also proposes revision to the Bases to reflect that the minimum

temperature for criticality provides assurance that the reactor is

operated within the bounds of the safety analyses. In addition, the

proposed application also includes an administrative change to correct

some typographical errors on page 3.1-25 of the Technical

Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) 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. The probability or the consequences of an

accident previously evaluated will not be affected because the

proposed changes will make the Minimum Temperature for Criticality

Specification (540 deg. F) more restrictive than the current

specification which allows reactor criticality at a temperature as

low as 450 deg. F. The proposed changes will also make the minimum

temperature for criticality consistent with the licensing basis

safety analyses. In addition, critical operation at Tavg less

than 540 deg. F will require operator response to restore Tavg

to [greater than or equal to] 540 deg. F within 15 minutes or be in

hot shutdown within the following 15 minutes. [***] [T]he minimum

temperature for criticality when defined to be within 7 deg. F below

the no-load Tavg value of 547 deg. F does not adversely affect

pressurizer operability, reactor vessel nil-ductility temperature,

the reactor protection system operability, nor the plant design

basis analyses and is supported by the current licensing basis

safety analyses. The presence of two separate alarms, each

annunciating on a 1-out-of-4 Tavg signal, will provide

assurance that constant Tavg monitoring is available during

approaches to criticality. The proposed change also increases the

surveillance frequency for Tavg instrument check when the

Tavg deviation and low Tavg alarms are not reset and the

control banks are above zero steps. Therefore, the proposed changes

have no effect on the probability or consequences of an accident

previously evaluated.

(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 changes do not involve the addition of any

new or different type of equipment, nor do they involve the

operation of equipment required for safe operation of the facility

in a manner different from those addressed in the Final Safety

Analysis Report. The safety analyses, which assume a critical

temperature of 547 deg. F, are applicable for critical temperatures

as low as 540 deg. F. The proposed changes will ensure that the

plant parameters are within their analyzed ranges and will increase

the surveillance frequency for the Tavg instrument check when

the Tavg deviation and low Tavg alarms are not reset and

the control banks are above zero steps. Therefore, the proposed

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

of accident from any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

:Response:

The proposed license amendment does not involve a significant

reduction in a margin of safety. The proposed changes do not affect

any safety related system or component operation or operability,

instrument operation, or safety system setpoints and do not result

in increased severity of any of the accidents considered in the

safety analyses. Operator response to a drop in temperature after

reaching criticality for a specified period of time will place the

reactor in the hot shutdown condition where the LCO [limiting

condition for operation] does not apply. The proposed changes are

being made to make the Technical Specifications consistent with the

licensing basis safety analyses and increase the surveillance

frequency. These changes have no effect on any margin of safety and,

therefore, do not create 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: Michael L. Boyle, Acting

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 31, 1994

Description of amendment request: This amendment request would

revise the diesel fuel oil storage operability requirements and the 5

minute diesel hot restart test conditions. In addition, the amendment

request also revises the Surveillance Requirements to allow the 24-hour

diesel generator endurance test to be conducted during any operational

condition.

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 addition of a 48 hour period to complete restoration of the

required fuel oil level prior to declaring the diesel generator

inoperable does not significantly increase the probability or

consequences of an accident previously evaluated. PSE&G believes

that the attendant risk of maintaining the diesel generator OPERABLE

status under temporary conditions where fuel oil supply is below

48,800 gallons (but greater than 44,709 gallons) is less than the

attendant risk of initiation and completion of shutdown actions

currently required by Technical Specifications under these

conditions. Since a minimum 44,709 gallon [6 day] supply of oil will

be maintained for these 48 hours, and procedures are implemented to

obtain replenishment fuel oil when the level falls below 48,800

gallons of fuel, and the probability of an event requiring the

onsite power sources during this brief period are low (as stated in

NUREG-1433), PSE&G concludes that this change does not increase the

likelihood of accidents occurring nor significantly affect the

performance of any system involved in the occurrence or mitigation

of the accidents.

The proposed amendment to allow the 24 hour diesel generator

endurance run to be conducted during any mode of operation does not

significantly increase the probability or consequences of an

accident previously evaluated in Chapter 15 of the UFSAR since the

capability to safely shutdown the plant following a Loss of Offsite

Power (LOP), LOCA or LOCA/LOP coincident with a single failure is

maintained throughout the surveillance test. The 24 hour endurance

test does not disable any of the automatic actuations and interlocks

of the diesel generator control functions, nor prevent the

satisfactory completion of the LOP or LOCA/LOP loading sequence if a

LOP or LOCA signal is received at any time during the test. Required

Class-1E onsite power OPERABILITY during normal operation, shutdown

cooling, loss of off-site power, and accident conditions will be the

same.

In addition, the performance of proposed Surveillance

Requirement 4.8.1.1.2.k.1 during Operational Conditions 1 or 2 will

not significantly increase the consequences of perturbations to any

of the electrical distribution systems that could result in a

challenge to steady state operation or to plant safety systems.

Performance of proposed Surveillance Requirement 4.8.1.1.2.k.1

during Operational Conditions 1 or 2, or failure of the

surveillance, will not cause, or result in, an anticipated

operational occurrence with attendant challenges to plant safety

systems that has not been previously analyzed for the existing

monthly surveillances.

Therefore, PSE&G concludes that this above change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The establishment of the new Surveillance Requirement

4.8.1.1.2.k.2, allows the five minute diesel hot restart test to be

performed at times other than after the 24 hour endurance run test,

but does not alter nor modify the test requirements currently

required by Surveillance Requirement 4.8.1.1.2.h.4.b. The proposed

Surveillance Requirement 4.8.1.1.2.k.2 maintains the test conditions

required by current Specifications, and satisfies the intent of

Regulatory Guide 1.9, Rev. 3, paragraph 2.2.10. Therefore, PSE&G

believes that this change does not involve a 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.

The proposed amendment does not involve any physical changes to

plant structures, systems or components, or change the manner in

which the plant is operated. Therefore, the proposed changes will

not increase the probability of accidents of a different type, nor

will they create malfunctions of a different type than any

previously evaluated in the SAR.

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

safety. The basis for this statement is outlined in Item 1 above.

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: Charles L. Miller

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 31, 1994

Description of amendment request: The amendment request would

establish an allowed out-of-service time (AOT) of 72 hours for any one

low pressure Emergency Core Cooling System injection subsystem

inoperable in addition to an inoperable High Pressure Coolant Injection

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 not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed change establishes a 72 hour Allowed Out-of-Service

Time (AOT) for coincidental High Pressure Coolant Injection system

(HPCI) and one low pressure injection/spray subsystem inoperability,

in accordance with NUREG-1433. In addition, PSE&G will justify

retaining the existing HCGS AOTs for one, two, and three Low

Pressure Coolant Injection (LPCI) subsystems inoperable.

The HCGS electrical distribution system that supplies the ECCS

(Emergency Core Cooling System) equipment with power contains four

(4) Emergency Diesel Generator (EDGs) while the electrical

distribution system of the NUREG-1433 plant contains only three (3)

EDGs. This additional, redundant system adds reliability to the

overall electrical distribution system. In NUREG-1433, the loss of

one EDG (Emergency Diesel Generator) results in the inoperability of

two (2) Low Pressure Injection (LPCI) pumps or one (1) LPCI and one

(1) 100% capacity CS subsystem, whereas the loss of one EDG at HCGS

would only result in the inoperability of one LPCI pump and one-half

of a CS subsystem (50% capacity). Another design feature of the HCGS

LPCI system that makes it more reliable than the plant discussed in

NUREG-1433 is that the LPCI system injects directly into the reactor

vessel via water boxes mounted inside the core shroud. The LPCI

system of the NUREG-1433 plant injects into the reactor vessel via

the recirculation loops, 2 LPCI pumps inject into each recirculation

loop. Therefore, following a rupture of a

recirculation line, a Design Base Accident Loss of Coolant

Accident, the NUREG-1433 plant would lose 2 LPCI subsystems, while HCGS

would maintain all 4 LPCI subsystems available for injection. The end

result is that, following assumed loss of a single EDG electrical

system, the HCGS plant is left with much more injection capability than

the NUREG-1433 plant. For these reasons:

*the HCGS AOT for one LPCI subsystem inoperable, given at least

one CS subsystem is operable, is 30 days as opposed to 7 days as

specified in NUREG-1433, and

*the HCGS AOT for two LPCI subsystems inoperable, given at least

one CS subsystem is operable, is 7 days. This AOT is the same as the

AOT in NUREG-1433 for one (1) LPCI subsystem inoperable. However,

this AOT is less restrictive than the action required in NUREG-1433

for one (1) LPCI subsystem inoperable with one (1) CS subsystem

inoperable, which is immediate entry into Technical Specification

3.0.3., and

*the HCGS AOT for three LPCI subsystems inoperable, given both

CS subsystems are operable, is 72 hours as opposed to immediate

entry into Technical Specification 3.0.3 as specified in NUREG-1433.

NUREG-1433 would be overly conservative for HCGS, since HCGS has

four (4) EDGs and four (4) physically separate channels of

electrical power, while the NUREG-1433 plant only has 3 EDGs. HCGS

has greater capability to supply power to its ECCS in an emergency

than the NUREG-1433 plant. Similarly, the incorporation of an AOT of 72

hours for HPCI and one low pressure injection/spray subsystem as

proposed in NUREG-1433 would also be conservative for HCGS.

In the HCGS safety analysis, the small break LOCA is the design

basis event for the HPCI system. The most limiting single failure

coincident with a large or small break LOCA is the failure of the DC

source (Channel A) common to the HPCI system, one CS subsystem, and

one LPCI subsystem. The change made by this submittal would

establish a 72 hour AOT for either one LPCI subsystem or one CS

subsystem to be inoperable coincident with HPCI being inoperable. As

discussed above, if an initiating event occurs while in the proposed

AOT for the ECCS, the resulting scenario would be within the design

basis of the HCGS. Even with a failure of one CS subsystem while

HPCI and LPCI are in a 72 hour LCO or a failure of one LPCI

subsystem while HPCI and CS are in a 72 hour LCO, HCGS would still

be within its design basis if an initiating event occurred. Adequate

core cooling is ensured by the operability of the ADS and the

remaining low pressure injection/spray subsystems.

The loss of feedwater flow transient analyzed in Chapter 15 of

the HCGS UFSAR assumes both the HPCI and RCIC systems to be

operable. The existing TS for HPCI allows the system to be

inoperable for up to 14 days. Sufficient injection capability is

assured because the RCIC system will be required to be operable if

the HPCI is inoperable. For this transient, the proposed 72 hour AOT

for the HPCI and one low pressure injection/spray subsystem

inoperable presents no challenge greater than the existing 14 day

HPCI AOT.

For the steam line breaks outside of containment, the HPCI

system is assumed unavailable. For feed line breaks outside of

containment, either the HPCI or RCIC systems are capable of

providing adequate cooling to the vessel to prevent cladding damage.

The RCIC system will be required to be operable during the proposed

72 hour AOT.

The Core Damage Frequency (CDF) at HCGS is calculated based on

Probabilistic Risk Assessment (PRA) models using present Technical

Specifications that do not allow simultaneous outages of either HPCI

and one LPCI subsystem or HPCI and one CS subsystem. The proposed

action statement, which would allow 72 hours of plant operation with

the HPCI system inoperable coincident with one low pressure ECCS

injection/spray subsystem (LPCI or CS) inoperable, was determined to

have a potential impact on 28 of the core damage sequences. These 28

core damage sequences were requantified with a new model (based on

the proposed action statement that would allow the simultaneous

outage), and none of the frequencies of the 28 sequences were

significantly affected.

In addition, the probability of an accident is not affected

because no physical modifications are being made to the plant.

Finally, implementation of this proposed change would reduce

unnecessary plant shutdowns without commensurate effects on safety

and minimize associated challenges to safety systems.

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

accident from any accident previously evaluated.

Establishing an AOT of 72 hours for the HPCI system inoperable

in addition to any one (1) low pressure injection/spray subsystem

does not alter the function of the equipment nor involve any type of

plant modification. Additionally, no new modes of plant operation

are involved with these changes. The proposed change therefore 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.

The basis for this statement is outlined in item 1 above.

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: Charles L. Miller

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 25, 1994

Description of amendment request: The proposed amendment would

change the minimum level in the Emergency Diesel Generator (EDG) Fuel

Oil Day Tank from 200 gallons to 360 gallons. The change is necessary

to support a required run time of 55-60 minutes from the day tank. The

Bases would also be changed to include reference to Amendment 59, which

authorized a change to the EDG surveillance requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The operation of the Hope Creek Generating Station (HCGS) in

accordance with the proposed change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment to the Technical Specifications will

bring the minimum allowable EDG fuel oil day tank levels into

agreement with the levels corresponding to EDG run times reviewed

and approved by the NRC in Amendment 59 to the Hope Creek Generating

Station Facility Operating License. Increasing the minimum level

requirement is conservative and when compared to the current level

requirement, will act to reduce the consequences of any accident or

malfunction of equipment important to safety previously evaluated in

the Updated Final Safety Analysis Report (FSAR). The proposed change

will have no impact on the probability of any accident.

2. The operation of the Hope Creek Generating Station (HCGS) in

accordance with the proposed change will not create the possibility

of a new or different kind of accident from any previously

evaluated.

The physical changes to the plant and to the manner in which the

plant is operated in the proposed revisions add conservatism to the

current requirements in the Technical Specifications. No new or

different kind of accident is created by the proposed change.

3. The operation of the Hope Creek Generating Station (HCGS) in

accordance with the proposed change does not involve a significant

reduction in a margin of safety.

The proposed revision, by maintaining a higher (more

conservative) Emergency Diesel Generator Day Tank level, will result

in an increase in margins 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: Charles L. Miller

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 25, 1994

Description of amendment request: The proposed amendment would

eliminate the requirement from the Hope Creek technical specifications

to perform Type C leak rate tests in accordance with 10 CFR 50,

Appendix J, of identified containment isolation valves that penetrate

the primary containment and terminate below the minimum water level in

the suppression chamber (torus). The valves would still be subject to

testing in accordance with the American Society of Mechanical Engineers

Boiler and Pressure Vessel Code and Type A leak rate tests (Integrated

Leak Rate Test) of the primary containment.

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 Containment Isolation Valves (CIVs) for which Appendix J,

Type C leak rate testing will no longer be performed are all on

lines which penetrate the Torus and terminate below the Torus

minimum water level. Since the Torus is designed and operated to be

filled with water during and following any postulated Design Basis

Accident, the CIVs will remain water sealed during these conditions.

Type C testing of individual CIVs per the requirements of

10CFR50 Appendix J is not necessary since no potential containment

bypass leakage path exists due to the water seal and closed system

piping. The CIVs will, however, continue to be tested pursuant to

the applicable requirements of Section XI of the American Society of

Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code.

In addition, the elimination of 10CFR50 Appendix J, Type C

testing for the subject CIVs will not affect the overall leak-tight

integrity of the Primary Reactor Containment, which will be

demonstrated during 10CFR50 Appendix J Type A Integrated Leak Rate

Testing as described in the Hope Creek Generating Station Updated

Final Safety Analysis Report (HCGS UFSAR) Section 6.2.6.1.

Consequently, radiological releases and their consequences due

to leakage of the subject CIVs will be minimized and within the

existing plant licensing basis.

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

accident from any accident previously evaluated.

This proposal does not involve any hardware or logic changes,

nor does it alter the way in which any plant systems operate. Post-

accident Containment isolation features, boundaries and system

interfaces are not affected by the changes. Therefore, there are no

new possibilities or types of accidents considered.

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

safety.

The proposed elimination of 10CFR50 Appendix J, Type C testing

for certain CIVs in lines which penetrate the Torus and terminate

below the minimum water level will not adversely affect the margins

of safety associated with the plant's licensing bases.

The water seal provided by the Torus, in conjunction with the

closed system piping, precludes post-accident bypass leakage.

Individual CIVs will be tested in accordance with the ASME B&PV Code

Section XI - Division 1, Article IWV-3000, as required. 10CFR50

Appendix J, Type A testing will ensure that the overall Containment

leakage rate is consistent with the plant's licensing bases. In

addition, any leakage associated with the subject CIVs will have

little radiological significance since it will involve liquid

releases (i.e., minimum fission products). CIV seat leakage will be

confined within the closed system piping downstream of the CIVs.

Existing Containment isolation features, boundaries and system

interfaces are not affected by the changes.

Since the CIVs and the systems they serve are all located in the

Reactor Building, any leakage (e.g., packing gland leakage) which

escapes the confines of the closed system piping, will be contained

within the Reactor Building. The Reactor Building is a

radiologically controlled area which is served by the Filtration,

Recirculation, and Ventilation System. This assures that all

radioactive releases to the environment are within the existing

plant licensing bases. The elimination of Type C testing for the

subject isolation valves will not affect the existing radiological

release evaluations currently described in the HCGS-UFSAR.

The proposed Amendment will not affect the functional capability

of any plant safety-related structures, systems or components nor

will it result in any relaxation of existing plant licensing bases.

In addition, the implementation of the proposed Amendment will

result in a reduction of radiological exposure to plant personnel.

Therefore, the proposed revision will not reduce a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: Charles L. Miller

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

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

County, New Jersey

Date of amendment request: March 28, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 3/4.3.2, Table 3.3-4, Item 7b, Sustained

Degraded Voltage for Salem Generating Station Units 1 and 2. The

amendment proposes to change the Sustained Degraded Voltage Trip

Setpoint from greater than or equal to 91.6% of bus voltage for less

than or equal to 13 seconds to greater than or equal to 94.6% of bus

voltage for less than or equal to 13 seconds, and the Allowable Value

for Sustained Degraded Voltage from greater than or equal to 91% of bus

voltage for less than or equal to 15 seconds to greater than or equal

to 94% of bus voltage for less than or equal to 15 seconds.

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 or malfunction of equipment important to

safety previously evaluated.

The proposed amendment will increase the minimum voltage

available at the vital buses and maintain vital loads within their

voltage ratings under degraded voltage conditions.

The change to the Sustained Degraded Voltage Trip Setpoint also

considered the minimum bus recovery voltage following a transient

that would reset the undervoltage relay to prevent unnecessary

transfers to the emergency diesel generators. Therefore, the

proposed amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated in

the UFSAR.

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

accident from any previously evaluated.

The proposed amendment to the Sustained Degraded Voltage Trip

Setpoint ensures a higher minimum voltage is available to all vital

loads during any electrical transient, and is sufficient to operate

these loads within their voltage rating. Therefore, the proposed

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

of accident from any previously evaluated.

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

safety.

The proposed amendment will increase the minimum voltage

available at the vital buses and maintain vital loads within their

voltage ratings. This ensures that the minimum voltage for any vital

load will continue to be available including under degraded voltage

conditions. Calculations have determined that a Sustained Degraded

Voltage Trip Setpoint of greater than or equal to 94.6% of bus

voltage for less than or equal to 13 seconds is the optimum setpoint

with an Allowable Value of greater than or equal to 94% of bus

voltage for less than or equal to 15 seconds to meet the current

design requirements. These values provide sufficient margin based on

the minimum voltage of 93.2% required to ensure that vital loads

will operate within their designed voltage range. Therefore, the

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 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: Charles L. Miller

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

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

Docket No. 50-354, Hope Creek Generating Station, Salem County, New

Jersey

Date of amendment request: April 13, 1994

Description of amendment request: The proposed changes revise the

Quality Assurance audit frequencies in the Hope Creek and Salem Unit

Nos. 1 and 2 Technical Specifications. These revisions will permit a

biennial audit frequency with a 25% extension to the frequency and

transfer subsequent control over the audit program to the Updated Final

Safety Analysis Report. Basis for proposed no significant hazards

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

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

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

consequences of an accident previously evaluated.

The likelihood that an accident will occur is neither increased

or decreased by the proposed Technical Specification changes which

affect review and audit frequencies. The proposed changes will not

impact the function or method of operation of plant equipment. Thus,

there is not a significant increase in the probability of a

previously analyzed accident due to the changes. Also, the

consequences of a malfunction of equipment important to safety

previously evaluated in the UFSAR is not increased by the changes.

The proposed changes affect review and audit frequencies. As

such, the proposed changes have no impact on accident initiators or

plant equipment, and thus, do not affect the probabilities or

consequences of an accident.

The audit program verifies that functions and methods of plant

operation have not been altered or degraded. This verification

process will continue, but on a more flexible basis. The proposed

changes will not alter the function or diminish the quality of the

audits. Frequency may only be decreased for those programs which

demonstrate acceptable performance. The changes provide added

flexibility in scheduling audits and facilitate resource allocation

to areas with perceived weaknesses. In this regard, the proposed

changes offer an opportunity for possible decreases in the

likelihood that an accident would occur.

Therefore, we conclude that the proposed changes do not

significantly increase the probabilities or consequences of an

accident.

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

accident from any accident previously evaluated.

The proposed changes do not involve changes to the physical

plant or operations. Since program audits do not contribute to

accident initiation, changes related to audit functions cannot

produce a new accident scenario or produce a new type of equipment

malfunction. Also, the [changes do] not alter any existing accident

scenarios. The proposed changes do not affect equipment or its

operation, and, thus, do not create the possibility of a new or

different kind of accident.

The proposed changes revise the audit frequency to biennial and

permit a 25% extension to this frequency. However, audit frequency

may only be decreased for those programs exhibiting acceptable

performance. The 25% extension is a provision to provide added

flexibility in scheduling audits and is not applicable to the audits

of the Facility Emergency and Security Plans.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident.

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

safety.

The proposed changes concerning conduct of reviews and audits do

not directly affect plant equipment or operation. Safety limits and

limiting safety system settings are not affected.

The proposed changes will not alter the function or diminish the

quality of the reviews and audits. For the audit program, the

changes propose a biennial audit frequency and a subsequent transfer

of control over the audit program to the Updated Final Safety

Analysis Report. This will not result in a loss of regulatory

control. 10 CFR 50.54(a)(3) requires that changes to the quality

assurance description report which reduce commitments be submitted

to the NRC prior to implementation. In addition, the audit frequency

may only be decreased for those programs which demonstrate

acceptable performance. The changes provide added flexibility in

scheduling audits and facilitate resource allocation to areas with

perceived weaknesses. The proposed changes should result in a more

effective audit program that will contribute to an improvement in

plant safety.

Therefore, use of the proposed Technical Specification changes

would not involve any 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079 (Salem Generating Station, Units

1 and 2) and Pennsville Public library, 190 S. Broadway, Pennsville,

New Jersey 08070 (Hope Creek Generating Station)

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

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

NRC Project Director: Charles L. Miller

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

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

County, New Jersey

Date of amendment request: May 3, 1994

Description of amendment request: The proposed amendments would

modify the Technical Specification (TS) for Combustible Gas Control (3/

4.6.4.1) by changing the surveillance frequency for performing the

channel functional test to once-per-quarter and the channel calibration

to once-per-refueling. Also, the TS for the Auxiliary Feedwater System

(3/4.7.1.2) would be changed to reduce the surveillance frequency for

performing pump operability tests to once-per-quarter on a staggered

test basis. These changes are consistent with the provisions of Generic

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

Reduce Surveillance Requirements For Testing During Power Operation.''

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 analyzed.

The proposed changes to the surveillance requirements for the

hydrogen analyzers and the Auxiliary Feedwater pumps are consistent

with the intent of Generic Letter 93-05, Line-Item Technical

Specification Improvements to Reduce Surveillance Requirements for

Testing During Power Operation, and NUREG-1366, Improvements to

Technical Specification Surveillance Requirements. The proposed

changes will modify surveillance frequency for both the hydrogen

analyzers and the Auxiliary Feedwater Pumps. Changing the

surveillance frequency for the hydrogen analyzers and the Auxiliary

Feedwater pumps does not affect the probability of occurrence or the

consequences of accidents identified in the UFSAR. No accident

precursors are being generated by the proposed increase in

surveillance frequency. Therefore, the proposed changes do not

involve a significant increase in the probability or consequences of

a previously analyzed accident.

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

accident.

The proposed changes to the surveillance requirements for the

hydrogen analyzers and the Auxiliary Feedwater pumps are consistent

with Generic Letter 93-05 and NUREG-1366. There are no modifications

or changes in operating conditions associated with the proposed

changes. Therefore, the proposed changes will not increase the

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

previously identified.

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

The Technical Specification operability requirements for the

hydrogen analyzers and the Auxiliary Feedwater pumps are not being

changed. Surveillance testing will still be performed on a routine

frequency. The proposed frequency will be capable of performing its

intended function and ensuring a consistent degree of reliability.

Therefore, the changes to the surveillance frequencies do not

involve a significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: Charles L. Miller

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

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

Alabama

Date of amendment request: September 29, 1993 (TS 333)

Description of amendment request: The proposed amendment would

increase the amount of Boron-10 required to be available in the Standby

Liquid Control System (SLCS).

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 any accident

previously evaluated.

The Standby Liquid Control System (SLCS) is designed to make the

reactor subcritical from rated power to a cold shutdown at any time

in core life with the control rods remaining withdrawn in the rated

power pattern. The increase in the amount of Boron-10 that is

required to be stored in the SLCS Solution Tank does not affect the

precursors for any accident or transient analyzed in Chapter 14 of

the BFN Final Safety Analysis Report (FSAR). Since there is no

change to an accident precursor, there is no increase in the

probability of any accident previously evaluated.

The increase in the amount of Boron-10 required to be stored in

the SLCS Solution Tank restores the ability of the SLCS to maintain

the Boron concentration required to ensure cold shutdown for future

anticipated core configurations. The minimum concentration of

natural Boron in the reactor vessel required to achieve cold

shutdown for the currently anticipated future core configurations is

660 ppm. The 660 ppm concentration was recommended by General

Electric in Service Information Letter (SIL) No. 325, June 1980.

The increase in the amount of Boron-10 required to be stored in

the SLCS Solution Tank restores the ability of the SLCS to maintain

the Boron concentration required to ensure cold shutdown for future

anticipated core configurations. The increased Boron-10 storage

requirements does not represent a change to the designed Boron

concentration capability of the SLCS. Since this change will ensure

the ability of the SLCS to mitigate the consequences of an accident

for future anticipated core designs, the change does not involve a

significant increase in the 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.

The increase in the amount of Boron-10 required to be stored in

the SLCS Solution Tank does not affect the function or operation of

any other system. The proposed change does not introduce any new

modes of operation or modify existing equipment design. Therefore,

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

different kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The SLCS shutdown margin is determined for each core

configuration by using the BWR simulator code to calculate the core

multiplication for the cold, xenon-free, all-rods-out condition at

the exposure point of maximum cold reactivity. The resulting k-

effective is subtracted from the critical k-effective of 1.0 to

obtain the SLCS shutdown margin. Increasing the amount of Boron-10

stored in the SLCS Solution Tank increases the SLCS shutdown margin.

Therefore, the proposed amendment results in an increased margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Alabama

Date of amendment request: September 30, 1993 (TS 312)

Description of amendment request: The proposed amendment removes

Technical Specification requirements for scram discharge volume air

header pressure trip switches in the Reactor Protection System (RPS)

for the Browns Ferry Nuclear Plant (BFN) Unit 2. The amendment also

clarifies the description of the RPS scram discharge volume high water

level bypass in the Technical Specifications for BFN Units 1, 2, and 3.

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 any accident

previously evaluated.

The Unit 2 scram pilot air header pressure trip switches perform

the same function as the high water level switches in the scram

discharge instrument volume. They automatically initiate control rod

insertion (scram) in the event that degraded air conditions are

detected in the BWR control air supply system. Since the scram pilot

air header pressure trip function is to ensure that the Control Rod

Drive (CRD) System is available to mitigate the consequence of an

accident or transient, and the removal of the scram pilot air header

pressure e trip switches scram function does not effect the

precursors for any accident or transient analyzed in Chapter 14 of

the BFN Final Safety Analysis Report (FSAR), there is no increase in

the probability of any accident previously evaluated.

The scram system has been analyzed, based on a plant-specific

maximum inleakage, and the removal of the scram pilot air header

pressure trip switches scram function would still result in a

successful scram, provided that the CRD leakage rate and the water

level instrumentation response characteristics remain within the

success criteria region. Administrative controls and periodic

measurements of the CRD and scram discharge system performance

ensure acceptable response time delays and stall flow rates. Since

the scram function would be successfully performed, the removal of

the scram pilot air header pressure trip switches scram function

does not involve a significant increase in the consequences of any

accident previously evaluated.

The clarification of the description of the scram discharge

volume high water level bypass in the Reactor Protection System does

not reflect a modification to plant equipment, maintenance

activities, or operating instructions. The revised description does

not effect the precursors for any accident or transient analyzed in

Chapter 14 of the BFN FSAR or equipment used in the mitigation of

these accidents or transients. Therefore, there is no increase in

the probability of any accident previously evaluated nor an increase

in the 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.

The design criteria for the scram system is contained in the

generic Safety Evaluation Report (SER), which was transmitted by NRC

letter to All BWR Licensees, dated December 9, 1980, BWR Scram

Discharge System. Section 4.2.4.1 of the generic SER states that the

CRD system shall be analyzed based on a plant-specific maximum

inleakage to ensure that the system function is not lost prior to

initiation of automatic scram. This analysis has been performed and

it was concluded that the removal of the scram pilot air header

pressure trip switches scram function would still allow a successful

scram, provided that the CRD leakage rate and the water level

instrumentation response characteristics remain within the success

criteria region. Administrative controls and periodic measurements

of the CRD and scram discharge system performance ensure acceptable

response time delays and stall flow rates.

The overall scram system design is in conformance with the

generic SER. No new system failure modes are created as a result of

removing the scram pilot air header pressure trip switches scram

function since, assuming the maximum CRD inleakage due to a degraded

air supply system, redundant and diverse level switches will

initiate a successful scram prior to this function becoming

disabled. The removal of the scram pilot air header pressure trip

switches scram function does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The clarification of the description of the scram discharge

volume high water level bypass in the Reactor Protection System does

not reflect a modification to plant equipment, maintenance

activities, or operating instructions. No new external threats,

system interactions, release pathways, or equipment failure modes

are created. Therefore, the clarification of this description 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.

The water level in the scram discharge instrument volume is

monitored by redundant Magnetrol and heated reference resistive

temperature devices (RTD) level switches. Since the level switches

are redundant and diverse, redundancy and diversity in the

instrumentation that initiates the scram signal is maintained even

with the removal of the scram pilot air header pressure trip

switches scram function.

The scram system has been analyzed, based on a plant-specific

maximum inleakage, and the removal of the scram pilot air header

pressure trip switches scram function would still result in a

successful scram, provided that the CRD leakage rate and the water

level instrumentation response characteristics remain within the

success criteria region. Since the scram system would successfully

operate to mitigate the consequences of accidents and transients

previously analyzed, the proposed amendment does not involve a

significant reduction in the margin of safety.

The clarification of the description of the scram discharge

volume high water level bypass in the Reactor Protection System does

not reflect a modification to plant equipment, maintenance

activities, or operating instructions. There is no change to the

licensing or design basis of the Reactor Protection System.

Therefore, the revised description does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and

50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone

County, Alabama

Date of amendment request: April 4, 1994 (TS322, Revision 1)

Description of amendment request: The proposed amendment supersedes

an amendment request dated March 25, 1993. Similar to the March 25,

1993, proposal, the revised amendment request for the Browns Ferry

Nuclear Plant (BFN), Units 1, 2, and 3, continues to seek elimination

of Technical Specifications (TS) for the reactor scram and main steam

line isolation functions associated with the main steam line radiation

monitors (MSLRMs). Additionally the revised amendment request proposes

elimination of TS for the remaining isolation functions associated with

the MSLRMs as well. The remaining isolation functions include main

steam line drain valve closure, reactor recirculation loop sample valve

closure, and main condenser mechanical vacuum pump (MVP) trip and

isolation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The probability of occurrence of previously evaluated accidents

is based on initial conditions and assumptions which are not

dependent directly or indirectly on the functions of the Main Steam

Line Radiation Monitors (MSLRMs). Elimination of the MSLRM scram and

isolation functions does not affect the operation of the other

Reactor Protection System or Primary Containment Isolation System

functions required to mitigate a Rod Drop Accident (RDA). Also,

eliminating the MSLRM functions does not affect the Control Rod

Drive System and, hence, cannot increase the probability of RDA. The

proposed change does not involve any increase in the probability of

an [any] previously evaluated accident.

There is no significant increase in the consequences of any

previously evaluated accident. Elimination of the MSLRM reactor

scram and isolation functions could potentially increase the amount

of radioactivity released following an RDA. However, the potential

increase in release from those previously determined is not

significant. Both the NEDO-31400 analysis and the additional BFN

dose calculations determined that the resulting doses (See Table B

[of this request]) remain well below the limits of Standard Review

Plan (SRP) 15.4.9 and 10 CFR 100.

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

new or different kind of accident from any accident previously

evaluated.

This proposed change deletes the reactor scram and isolation

functions of the MSLRMs. The sole purpose of these functions is to

assist in mitigating the consequences of an RDA, a previously

analyzed event. This event is terminated by a high flux scram.

The NEDO-31400 RDA analysis without the MSLRM scram and MSIV

[Main Steam Isolation Valve] closure functions has been reviewed and

accepted by the NRC. An RDA without the MSLRM MVP isolation function

has been previously reviewed by the NRC as documented in Section

14.6.2 of the Updated Final Safety Analysis Report (UFSAR). An RDA

without the MSLRM recirculation sample line isolation coupled with a

sample line break is no different than that if the normal RWCU

[Reactor Water Cleanup] [system] sample path is in service (see

Section III.C) [of this request]. However, TVA has evaluated the

consequences of this event and determined the MSLRM sample line

isolation function is not required to maintain radioactive releases

within the acceptable limits.

The NEDO-31400 RDA analysis and the additional BFN offsite dose

calculations show that the elimination of the MSLRM functions does

not create the possibility of a new or different kind of

accident from any accident previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

A reliability assessment of the elimination of the MSLRM scram

function on reactivity control failure frequency and core damage

frequency was performed as part of the NEDO-31400 analysis. The

results of the NEDO-31400 analysis indicated a negligible increase

in reactivity control failure frequency associated with deletion of

the MSLRM scram function. However, this increase is offset by a

reduction in the frequency of transient initiating events

(inadvertent scrams). This reduction in transient initiating events

represents a net reduction in core damage frequency of 0.3 percent.

BFN TS Bases 3.2 states that the MSLRMs are provided to detect

gross fuel failures as in an RDA and provide main steam isolation

valve closure to maintain radiological releases below the 10 CFR 100

limits. The BFN offsite dose calculations and the NRC's Safety

Evaluation Report approving the NEDO-31400 RDA analysis document

that the potential radiological release consequences following an

RDA without the MSLRMs reactor scram and isolation functions are

still well within the 10 CFR 100 limits. Thus, this change will not

result in a significant 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: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: May 16, 1994 (TS 93-18)

Description of amendment request: The proposed change would revise

Surveillance Requirement 4.8.1.1.1.b to reflect installation of common

station service transformers with auto load tap changers and the

resulting change to the alignment of this portion of the distribution

system. The proposed amendment would replace the terms ``normal'' and

``alternate'' with ``unit generator supported'' and ``preferred power

(GDC 17),'' respectively, circuits that must be demonstrated operable.

Basis for proposed no significant hazards consideration

determination: Asrequired by 10 CFR 50.91(a), the licensee has provided

its analysis of the issue of no significant hazards consideration,

which is presented below:

TVA has evaluated the proposed technical specification (TS)

changeand has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

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

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The description clarification for the circuits in TS 4.8.1.1.1.b

does not change any TS requirement and will not alter plant

functions or components as a result of normal alignment to the

common station service transformers (CSSTs). By implementing normal

CSST alignment to unit power, SQN has reduced the required transfers

to obtain alignment to the preferred power circuit under accident

conditions. The verifications provided by TS Surveillance

4.8.1.1.1.b will continue to ensure the capability of this transfer

during occasional use of the unit station service transformer (USST)

for maintenance or repair activities. The accidents described in

Chapter 15 of the SQN Updated Final Safety Analysis Report assume

the unit power supply to be from the USST circuit that requires

transfer to the CSST after turbine trip. This analysis is bounding

for normal alignment to the CSST that does not require a transfer.

These changes minimize the impact of malfunctions because transfers

will not be required as often. The consequences of an accident are

therefore not increased by this change. The proposed unit power

alignment to the CSSTs has always been used for unit start-up; and

since no plant functions are changed as a result of normal alignment

to the CSSTs, other than a lower potential for transfers, the

probability of an accident is not increased.

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

from any previously analyzed.

The TS 4.8.1.1.1.b clarification does not change plant functions

and only provides a more appropriate description of the circuits

involved in the transfer function. The transfer from the USST

circuit to the CSST circuit will still be available when SQN is

using the USST circuit, but the need for these transfers will be

reduced by normal alignment of unit power to the preferred CSST

circuit. These changes will not create a new or different kind of

accident because the plant functions remain unchanged and only the

CSST alignment for normal unit power and circuit descriptions is

altered. This CSST alignment has always been used for unit start-up

and is being expanded to apply to normal power operation in addition

to start-up and accident conditions.

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

The clarification for the unit power circuit descriptions does

not alter the intent of requirements for this surveillance. Plant

functions and setpoints remain unchanged for automatic transfers

from the USST circuit to the CSST circuit. This circuit description

change provides a clearer explanation of how the transfer functions.

The margin of safety has not been altered for unit power alignment

to the USST or CSST, and this change will support SQN's use of the

CSST for the normal unit power circuit. Therefore, these changes do

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: December 6, 1993

Description of amendment request: The proposed amendment would

remove the requirement to perform jet pump integrity and operability

surveillances in the idle loop during operation with one recirculation

pump. It also provides consistency between the Bases and Section 4.6.F

on jet pump surveillance requirements, and includes some administrative

changes within the Bases.

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:

Removing the surveillance requirement to examine the jet pump

flow patterns within the idle loop during single loop operation does

not increase the probability or consequences of an accident

previously evaluated. The probability of loss of the ability of such

a jet pump to maintain a refloodable volume of two thirds core

height is not increased. The stresses undergone without forced

recirculation flow are very small and therefore a jet pump within an

idle recirc loop should not physically degrade during this time. The

ability of the core standby cooling systems to provide their

intended function is not degraded by removal of this surveillance

requirement; therefore the consequences of the recirculation line

break accident are not increased.

This change will remove a surveillance requirement that has been

determined to be unnecessary based upon the operating conditions

during single recirculation loop operation. Other surveillance

requirements within the technical specifications adequately

determine jet pump operability when stresses exist which could cause

a jet pump failure to occur. The change will not affect any plant

hardware, plant design, safety limit settings, or plant system

operation, and therefore does not modify or add any initiating

parameters that would significantly increase the probability or

consequences of any previously analyzed accident.

As discussed above, there is no increase in the probability of a

jet pump failing and being undetectable with the proposed technical

specification surveillances. Therefore, there is no increase in the

possibility of not being able to maintain a floodable volume of two

thirds core height during the design basis LOCA [loss-of-coolant

accident]. As a result, there is no 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: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301

Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,

One International Place, Boston, Massachusetts 02110-2624

NRC Project Director: Walter R. Butler

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.

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of application for amendment: January 14, 1994

Brief description of amendment request: The proposed amendment

would revise the technical specifications (TSs) by removing component

lists from the TSs in accordance with NRC Generic Letter (GL) 91-08 and

by removing the schedule for withdrawal of reactor vessel material

specimen capsules from the TSs in accordance with GL 91-01.

Date of individual notice in Federal Register: May 23, 1994 (59 FR

26675)

Expiration date of individual notice: June 22, 1994

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

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

Project No. 2 Benton County, Washington

Date of application for amendment: December 6, 1993, supplemented

by letter May 6, 1994

Brief description of amendment request: The proposed amendment

would revise Technical Specifications (TS) 3/4.4.2 and 3/4.5.1 to

require main steam system and automatic depressurization system safety/

relief valve (SRV) surveillance testing within 12 hours after steam

pressure and flow have been found to be adequate to do the testing.

Also, TS Table 4.3.7.5-1 would be revised to require SRV position

indicator surveillance testing within 12 hours after steam pressure and

flow have been found to be adequate to do the testing.

Date of individual notice in Federal Register: May 13, 1994 (59 FR

25131)

Expiration date of individual notice: June 13, 1994

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

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

Project No. 2 Benton County, Washington

Date of application for amendment: May 5, 1994

Brief description of amendment request: The proposed amendment

would change the plant operating license to rename three primary

containment isolation check valves listed in the technical

specifications. The licensee is making an administrative change to

rename valve PI-EFC-X29d to make its number consistent with other

similar valves in the technical specifications. The licensee is

renaming excess flow check valves PI-EFCX-72f and PI-EFCX-73e because

they are replacing them with swing check valves that have a different

numbering nomenclature.

Date of individual notice in Federal Register: May 12, 1994 (59 FR

24762)

Expiration date of individual notice: June 13, 1994

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

rooms for the particular facilities involved.

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

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

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

Date of application for amendments: January 20, 1994

Brief description of amendments: The amendments increase the

departure from nucleate boiling ratio limit from 1.24 to 1.30 to

accommodate the uncertainties in core inlet flow distribution. The

amendments also add the analytical method supplement entitled ``System

80TM Inlet Flow Distribution'' to the list of methods used to

determine core operating limits.

Date of issuance: May 26, 1994

Effective date: May 26, 1994

Amendment Nos.: 76, 62 and 48

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

amendments revised the Technical Specifications.

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

12358) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 26, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of application for amendments: April 1, 1993

Brief description of amendments: The amendments make

administrative, editorial, and format changes to the Operating

Licenses. These changes include the deletion or incorporation, as

appropriate, of all handwritten or ``pasted-up'' changes and the

removal of all previous license conditions that have been completed to

the satisfaction of the Commission. The changes and reformatting result

in the Operating Licenses containing only those license conditions that

are currently applicable.

Date of issuance: May 20, 1993

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

30 days.

Amendment Nos.: 189 and 166

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Facility Operating Licenses.

Date of initial notice in Federal Register: May 12, 1993 (58 FR

28052) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated May 20, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

Station, Plymouth County, Massachusetts

Date of application for amendment: February 9, 1994

Brief description of amendment: This amendment changes the

Technical Specifications and the License. These changes consist of

revised wording for the License, clarifying wording to aid operators in

selecting the correct pressure/temperature curve during startup and

shutdown operations, and removal of certain obsolete mechanical snubber

acceptance criteria.

Date of issuance: May 16, 1994

Effective date: May 16, 1994

Amendment No.: 153

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications and License.

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

12359) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 16, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: March 1, 1993

Brief description of amendment: The amendment revises the Technical

Specifications to provide specific actions to take if primary

containment leakage limits are exceeded and cannot be restored when

reactor coolant temperature is greater than 200 deg. F.

Date of issuance: May 26, 1994

Effective date: May 26, 1994

Amendment No. 60

Facility Operating License No. NPF-58. This amendment revised the

Technical Specifications.

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

14896) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 26, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

Neck Plant, Middlesex County, Connecticut

Date of application for amendment: January 17, 1994

Brief description of amendment: The amendment removes Technical

Specification 3/4.4.12, ``Failed Fuel Rods'' and its associated Bases

Section 3/4.4.12.

Date of Issuance: May 17, 1994Effective date: As of the date of

issuance to be implemented within 30 days.

Amendment No.: 171

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

Technical Specifications.

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

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

contained in a Safety Evaluation dated May 17, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: March 19, 1993

Brief description of amendment: The amendment changed Technical

Specification 4.18.6 and Table 4.18-2 to make the requirements for C-3

reports on steam generator tube inspections consistent with the Babcock

& Wilcox Standard Technical Specifications.Date of issuance: May 19,

1994Effective date: May 19, 1994

Amendment No.: 172

Facility Operating License No. DPR-51. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 21, 1993 (58 FR

39049) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 19, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

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

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

Date of application for amendments: September 3, 1993

Brief description of amendments: These amendments eliminate the

reliance on the five ``cranking'' diesel generators by replacing the

motor driver on one of the standby steam generator feedwater pumps with

a dedicated diesel driver.

Date of issuance: May 20, 1994

Effective date: May 20, 1994

Amendment Nos. 164 and 158Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 13, 1993 (58 FR

52985) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 20, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

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

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

Date of application for amendment: November 30, 1993

Brief description of amendment: The amendment revises the plant

Technical Specifications (TS) to remove the protective and maximum

allowable setpoint limits for axial power imbalance and the trip

setpoint for nuclear overpower based on reactor coolant system (RCS)

flow (flux-to-flow) from the TS and relocate them to the existing TMI-1

Core Operating Limits Report (COLR).

Date of Issuance: May 23, 1994

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

within 30 days.

Amendment No.: 184

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

Technical Specifications.

Date of initial notice in Federal Register: January 19, 1994 (59 FR

2867). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 23, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: March 3, 1994

Brief description of amendment: The amendment revises the technical

specifications in accordance with the guidance provided by Generic

Letter 93-08, ``Relocation of Technical Specification Tables of

Instrument Response Time Limits.''

Date of issuance: May 19, 1994

Effective date: May 19, 1994

Amendment No.: 73

Facility Operating License No. NPF-47. The amendment revised the

Technical Specifications.

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

12380) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 19, 1994.No significant

hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

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

Linn County, Iowa

Date of application for amendment: January 21, 1994

Brief description of amendment: The amendment revised the Technical

Specifications by changing the name of one operator and one of the

owners of the Duane Arnold Energy Center from Iowa Electric Light and

Power Company to IES Utilities Inc. The title of the position

responsible for the management of the Nuclear Division has also been

changed to Vice President, Nuclear from Manager-Nuclear Division. One

blank page was deleted and several other editorial changes were made.

Pages 1 through 4 of the Facility Operating License were revised to

reflect the corporate name change and a spelling error.

Date of issuance: May 13, 1994

Effective date: May 13, 1994

Amendment No.: 198

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

license and Technical Specifications.

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

10008) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 13, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

Linn County, Iowa

Date of application for amendment: December 22, 1993

Brief description of amendment: The amendment revised the Technical

Specifications by correcting several typographical and administrative

errors.

Date of issuance: May 18, 1994

Effective date: May 18, 1994

Amendment No.: 199

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

Technical Specifications.

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

10008) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 18, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

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

Connecticut

Date of application for amendment: April 14, 1994, as

supplementedApril 20, 1994.

Brief description of amendment: The amendment revises the Technical

Specifications (TS) to change the laboratory testing protocol for the

charcoal absorbers for the Control Room Emergency Ventilation System

(TS 3.7.6.1), the Enclosure Building Filtration System (TS 3.6.5.1) and

the Storage Pool Ventilation System (TS 3.9.15).

Date of issuance: May 23, 1994

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

within30 days.

Amendment No.: 175

Facility Operating License No. DPR-65. Amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes (59 FR 23085, May 4, 1994) That

notice provided an opportunity to submit comments on the Commission's

proposed no significant hazards consideration determination. No

comments have been received. The notice also provided for an

opportunity to request a hearing by June 3, 1994, but indicated that if

the Commission makes a final no significant hazards consideration

determination any such hearing would take place after issuance of the

amendment.The Commission's related evaluation of the amendment, finding

of exigent circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated May

23, 1994.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

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

County,Minnesota

Date of application for amendments: February 14, 1994

Brief description of amendments: The amendments revise Technical

Specifications to reflect the new configuration for the Unit 1 480V

safeguards bus arrangement (two 480V safeguards buses fed by each 4160V

safeguards bus). These changes make the specifications the same for

both units since the configuration for the two units will become the

same at the completion of the current outage.

Date of issuance: May 17, 1994

Effective date: May 17, 1994

Amendment Nos.: 110 & 103

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

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

14892) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 17, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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 application for amendment: January 6, 1993

Brief description of amendment: The amendment revised the Technical

Specifications (TS) to incorporate the following changes:

(1) The residual heat removal (RHR) pump flow calibration frequency

(specified in TS Table 4.1-1) was changed to accommodate operation on a

24-month cycle.

(2)The RHR loop isolation valve automatic isolation and interlock

testing frequency (specified in TS Table 4.1-3) was changed to

accommodate operation on a 24-month cycle.

(3)The RHR system leakage testing frequency (specified in TS

Section 4.4.I.4) was changed to accommodate operation on a 24-month

cycle.

(4)The recirculation pump testing frequency (specified in TS

Section 4.5.B.1.a) was changed to accommodate operation on a 24-month

cycle.

(5)The accumulator check valve operability testing frequency

(specified in TS Section 4.5.B.2.b) was changed to accommodate

operation on a 24-month cycle.

(6)The safety injection (SI)/RHR check valve gross leakage testing

frequency (specified in TS Section 4.5.B.2.c) was changed to

accommodate operation on a 24-month cycle.

These changes followed the guidance provided in Generic Letter 91-

04, ``Changes in Technical Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,'' as applicable.

In addition, the gross leakage surveillance requirements for

certain SI/RHR system check valves (specified in TS Section 4.5.B.2.d)

was changed to implement requirements as set forth in NRC generic

letter, dated February 23, 1980, regarding testing of low pressure

injection (LPI)/RHR check valves. Therefore, Item A.5 of the February

11, 1980, Confirmatory Order is considered rescinded.

Date of issuance: May 20, 1994

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

30 days.

Amendment No.: 148

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 17, 1993 (58

FR 8777) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 20, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: October 18, 1993, and suplement

dated March 7, 1994

Brief description of amendment: This amendment revised TS 3/4.3.2

and associated Bases to increase surveillance test intervals and add

out-of-service times for isolation actuation instrumentation. The

changes are in accordance with General Electric Company's Licensing

Topical Reports which have been previously reviewed and approved by the

NRC staff.

Date of issuance: May 25, 1994

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

within 60 days of the date of issuance.

Amendment No.: 70

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

Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64615) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: October 26, 1993 and

supplemented by letter dated December 14, 1993.

Brief description of amendment: The amendment request revised the

MINIMUM CHANNELS OPERABLE requirement for suppression pool water

temperature instruments, Accident Monitoring ACTION STATEMENTS, and

removes ACTIONS and Surveillance Requirements for suppression chamber

temperature and level instruments.

Date of issuance: May 25, 1994

Effective date: May 25, 1994

Amendment No.: 71

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

Technical Specifications.

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

FR 67861) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama.

Date of amendments request: October 14, 1993

Brief description of amendments: The requested changes incorporate

changes allowing longer surveillance test intervals (STIs) and allowed

outage times (AOTs) for the reactor trip system (RTS) and engineered

safety features actuation system (ESFAS) instrumentation into the

Technical Specifications. The proposed changes also revise certain RTS/

ESFAS functions, minimum channels operable, channel calibration, and

channel functional test requirements to ensure they are in concert with

the Westinghouse Standard Technical Specifications and WCAP-10271,

``Evaluation of Surveillance Frequencies and Out-of-Service Times for

Reactor Protection Instrumentation Systems.''

Date of issuance: May 16, 1994

Effective date: May 16, 1994

Amendment Nos.: 107 and 99

Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (57 FR

17605) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 16, 1994.No significant

hazards consideration comments received: No

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

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

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

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

Dates of application for amendments: April 6, 1992 (TS 308), and

September 28, 1992 (TS 326)

Brief description of amendments: The amendments change the

Technical Specifications (TS) to add Automatic Depressurization System

(ADS) high drywell pressure bypass timer requirements, revise the ADS

timer trip level setting, increase the number of ADS valves required to

be operable for startup, and revise the limiting conditions for

operation with inoperable ADS valves. The ADS bases have also been

revised for consistency with these TS changes.

Date of issuance: May 19, 1994

Effective date: May 19, 1994

Amendment Nos.: 205 and 178

Facility Operating License Nos. DPR-33 and DPR-68: Amendments

revised the Technical Specifications.

Dates of initial notice in Federal Register: May 27, 1992 (57 FR

22269) and November 25, 1992 (57 FR 55593).The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

May 19, 1994.No significant hazards consideration comments received:

None

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

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

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

Date of application for amendments: February 9, 1994; supplemented

April 13, 1994 (TS 93-21)

Brief description of amendments: The amendments change Technical

Specification Table 3.3-11 to reflect the addition of ionization fire

detectors to Fire Zones 184, 185, 186, and 187.

Date of issuance: May 23, 1994

Effective date: May 23, 1994

Amendment Nos.: 181 and 173

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

revise the technical specifications.

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

12369) The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated May 23, 1994. No significant

hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

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

Date of application for amendments: October 1, 1993; supplemented

March 29, 1994 (TS 93-09)

Brief description of amendments: The amendments revise the

setpoints and time delays for the Auxiliary Feedwater loss of power and

6.9 kv shutdown board loss-of-voltage and degraded-voltage

instrumentation. In addition, the description, total number of

channels, channels to trip, minimum channels operable, actions, trip

setpoints, allowable values, channel checks, and channel functional

test requirements for loss-of-power instrumentation have been revised.

Date of issuance: May 24, 1994

Effective date: May 24, 1994

Amendment Nos.: 182 and 174

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

revise the technical specifications.

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

4947) The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated May 24, 1994.No significant

hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: March 30, 1994

Brief description of amendments: The proposed amendments revise the

technical specifications by adding an alternative method for verifying

that the emergency diesel generator fuel oil meets requirements.

Date of issuance: May 13, 1994

Effective date: May 13, 1994, to be implemented within 30 days of

issuance.

Amendment Nos.: Unit 1 - Amendment No. 24; Unit 2 - Amendment No.

10

Facility Operating License Nos. NPF-87 and NPF-89. The amendments

revised the Technical Specifications.

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

17606) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 13, 1994 No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: February 25, 1994

Brief description of amendments: The amendments revise the

surveillance frequency of the nozzles in the Quench Spray and

Recirculation Spray Systems from 5 to 10 years. The change is in

accordance with NRC Generic Letter 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements at

Power Operation,'' dated September 27, 1993.

Date of issuance: May 16, 1994

Effective date: May 16, 1994

Amendment Nos.: 182 and 163

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

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

17608) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 16, 1994No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: October 4, 1993

Brief description of amendments: These amendments revise the NA-1&2

Technical Specifications requirements to allow the use of ZIRLO

material for fuel cladding.

Date of issuance: May 26, 1994

Effective date: May 26, 1994

Amendment Nos.: 183 and 164

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57859) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 26, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

Date of application for amendments: February 25, 1994

Brief description of amendments: The amendments revise the

surveillance frequency of the nozzles in the containment spray and

recirculation spray systems from 5 to 10 years.

Date of issuance: May 20, 1994

Effective date: May 20, 1994

Amendment Nos. 191 and 191

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

revised the Technical Specifications.

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

17608) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 20, 1991.No significant

hazards consideration comments received: No

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,

and at the local public document room for the particular facility

involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By July 8, 1994, the licensee

may file a request for a hearing with respect to issuance of the

amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

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

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

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

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

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

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

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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

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

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

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

the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the

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

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

the Commission by a toll-free telephone call to Western Union at 1-

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

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

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

Commonwealth Edison Company, Docket No. STN 50-457, Braidwood

Station, Unit No. 2, Will County, Illinois

Date of application for amendment: April 21, 1994

Brief description of amendment: The amendment effects a one-time

only change to Technical Specification (TS) Surveillance Requirement

4.7.1.1 by add

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